Thermal shield support degradation in pressurized water reactors

International Nuclear Information System (INIS)

Sweeney, F.J.; Fry, D.N.

1986-01-01

Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

A Sensitivity Study on the Radiation Shield of KSPR Space Reactor

Energy Technology Data Exchange (ETDEWEB)

Cerba, S.; Lee, Hyun Chul; Lim, Hong Sik; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The idea of a space reactor was realised some decades ago and since that time several research activities have been performed into this field. The US National Aeronautics and Space Administration (NASA) has been developing a small fast reactor called as fission power system (FPS) for deep space mission, where highly enriched uranium (HEU) is used as fuel. On the other hand, other researchers have also surveyed a thermal reactor concept with low enriched uranium (LEU) for space applications. One of the main concerns in terms of a space reactor is the total size and the mass of the system including the reactor itself as well as the radiation shield. Since the reactor core is a source of neutrons and gamma photons of various energies, which may cause severe damage on the electronics of the space stations, the questions related to the development of a radiation shield should be address appropriately. The proposal of a radiation shield for a small space reactor is discussed in this paper. The requirements for the radiation shield have been addressed in terms of maximal absorbed doses and neutron flounces during 10 years of operation. In this study a radiation shield design for a small space reactor was investigated. All the presented calculations were performed using the multi-purpose stochastic MCNP code with temperature dependent continuous energy ENDF/B VII.0 neutron and photon cross section libraries. The aim of this study was to design a neutron and gamma shield that can meet the requirements of 250 Gy absorbed during 10 years of reactor operation. The comparison with a fast reactor design showed that high content of {sup 238}U strongly influences the shielding mass. This phenomenon is due to the higher photon production in case of the KSPR design and therefore the use of high {sup 235}U enrichments and the operation in fast neutron spectrum may be more desirable. In case if the KSPR space reactor the best shielding performance was achieved while utilizing a multi

Radiation shield for PWR reactors

International Nuclear Information System (INIS)

Esenov, Amra; Pustovgar, Andrey

2013-01-01

One of the chief structures of a reactor pit is a 'dry' shield. Setting up a 'dry' shield includes the technologically complex process of thermal processing of serpentinite concrete. Modern advances in the area of materials technology permit avoiding this complex and demanding procedure, and this significantly decreases the duration, labor intensity, and cost of setting it up. (orig.)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

McKissock, B.I.; Bloomfield, H.S.

1990-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

Mckissock, B.I.; Bloomfield, H.S.

1989-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

The optimum shielding for a power reactor using local components

International Nuclear Information System (INIS)

AlHajali, S.; Kharita, M. H.; Yousef, S.; Naoom, B.; Al-Nassar, M.

2009-07-01

Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

Radiological shielding of low power compact reactor: calculation and design

International Nuclear Information System (INIS)

Marino, Raul

2004-01-01

The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es

Estimation of temperature distribution in a reactor shield

International Nuclear Information System (INIS)

Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

1989-01-01

Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

International Nuclear Information System (INIS)

Atkins, D.F.; Lillie, A.F.

1983-01-01

As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

Tungsten-based composite materials for fusion reactor shields

International Nuclear Information System (INIS)

Greenspan, E.; Karni, Y.

1985-01-01

Composite tungsten-based materials were recently proposed for the heavy constituent of compact fusion reactor shields. These composite materials will enable the incorporation of tungsten - the most efficient nonfissionable inelastic scattering (as well as good neutron absorbing and very good photon attenuating) material - in the shield in a relatively cheap way and without introducing voids (so as to enable minimizing the shield thickness). It is proposed that these goals be achieved by bonding tungsten powder, which is significantly cheaper than high-density tungsten, with a material having the following properties: good shielding ability and relatively low cost and ease of fabrication. The purpose of this work is to study the effectiveness of the composite materials as a function of their composition, and to estimate the economic benefit that might be gained by the use of these materials. Two materials are being considered for the binder: copper, second to tungsten in its shielding ability, and iron (or stainless steel), the common fusion reactor shield heavy constituent

Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

Shield structure for a nuclear reactor

International Nuclear Information System (INIS)

Rouse, C.A.; Simnad, M.T.

1979-01-01

An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

International Nuclear Information System (INIS)

Berg, Thomas A.; Disney, Richard K.

2004-01-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs

Preliminary shielding design evaluation for reactor assembly of SMART

International Nuclear Information System (INIS)

Kim, Kyo Youn; Kang, Chang M.; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

1999-03-01

This report describes a preliminary evaluations of SMART shielding design near the reactor core by using the DORT two-dimensional discrete ordinates transport code. The results indicate that maximum neutron fluence at the bottom of reactor vessel is 1.64x10 17 n/cm 2 and that on the radial surface of reactor vessel is 6.71x10 16 n/cm 2 . These results meet the requirement, 1.0x10 20 n/cm 2 , in 10 CFR 50.61 and the integrity of SMART reactor vessel is confirmed during the lifetime of reactor. (Author). 20 refs., 11 tabs., 8 figs

Summary of the fifth international conference on reactor shielding

International Nuclear Information System (INIS)

Roussin, R.W.; Abbott, L.S.; Bartine, D.E.

1977-01-01

The Fifth International Conference on Reactor Shielding was held April 18-23, 1977 in Knoxville, Tennessee. The meeting was the largest in the series and attracted participants from 34 countries. The 10 invited papers and 10 of the contributed papers, selected as being representative of the Conference by the Technical Program Committee, are published in this issue of ATOMKERNENERGIE. This collection of papers demonstrates that the field of nuclear reactor shielding has developed into a mature discipline while retaining a definite vitality. (orig.) [de

Shield design and streaming calculations for the sodium cooled PEC reactor

International Nuclear Information System (INIS)

Prosperi, M.; Tavoni, R.; Travaglini, N.

1977-01-01

This paper summarises the shielding calculations carried out for the PEC reactor. A brief description of calculation methods and of the work carried out to set them up is given; the most representative calculations with the relative isoflux curves are also referred. A general outline is then given for the main shielding problems of the PEC reactor

Numerical simulation of a reinforced concrete shield around a nuclear reactor

International Nuclear Information System (INIS)

Mahama, Mumuni Salifu

1996-02-01

Ghana currently operates a Research Reactor and other nuclear facilities including a Gamma Irradiation Facility, a Radiographic Non-Destructive Testing laboratory and would be operating in the nearest future a Radiotherapy Centre. Each of these has a concrete radiation shield as a major safety device. In carrying out its functions, a concrete radiation shield may be subjected to thermal and mechanical stresses. A facility for analysing these stresses is desirable. Two computer codes have been developed under this programme for radiation shielding computation and stress analysis of cylindrical reactor shields. (au)

Up-dating of the RA-0 reactor shielding. Gamma and neutron isodoses

International Nuclear Information System (INIS)

Murua, Carlos A.; Chautemps, Norma A.; Ackerley, Alejandro F.; Alexeiew, Vladimiro

1999-01-01

A comparative analysis of the historical shielding configurations of the RA-0 reactor is performed and the comparison methodology is described. The gamma and neutron dose mapping of the last two stages of the reactor shielding has been carried out and the results are analysed

Preliminary shielding analysis of VHTR reactors

International Nuclear Information System (INIS)

Flaspoehler, Timothy M.; Petrovic, Bojan

2011-01-01

Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

Shielding considerations for advanced space nuclear reactor systems

International Nuclear Information System (INIS)

Angelo, J.P. Jr.; Buden, D.

1982-01-01

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

International Nuclear Information System (INIS)

Abulfaraj, W.H.; Kamal, S.M.

1994-01-01

The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concrete shielding. Multiattribute utility theory is selected to accommodate decision maker's preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Illmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy weight heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Illmenite Serpentine concrete. (Author)

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2016-11-15

Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

Radiation distribution through serpentine concrete using local materials and its application as a reactor biological shield

International Nuclear Information System (INIS)

Kansouh, W.A.

2012-01-01

Highlights: ► New serpentine concrete was made and examined as a reactor biological shield. ► Ilmenite–limonite concrete is a better reactor biological shield. ► New serpentine concrete is a better reactor fast neutrons shield than ordinary and hematite–serpentine concretes. ► Serpentine concrete has lower properties as a reactor total gamma rays shields. - Abstract: In the present work attempt has been made to estimate the shielding parameters of the new serpentine concrete (density = 2.4 g/cm 3 ) using local materials on the shielding parameters for two types of heat resistant concretes, namely hematite–serpentine (density = 2.5 g/cm 3 ) and ilmenite–limonite (density = 2.9 g/cm 3 ). Shielding parameters for ordinary concrete (density = 2.3 g/cm 3 ) were also discussed. These parameters were determined experimentally for serpentine concrete and compared with previously published values for other concretes, which had also been obtained using local materials. The leakage spectra of reactor fast neutrons and total gamma photon beams from cylindrical samples of these concrete shields were also investigated using a collimated beam from ET-RR-1 reactor. A neutron–gamma spectrometer was used in order to obtain pulse height spectra of reactor fast neutrons and the total gamma rays leakage through the investigated concrete samples. These spectra were utilized to obtain the energy spectra required in these investigations. Removal cross section Σ R (E n ) and linear attenuation coefficient μ(E g ) for reactor fast neutrons and total gamma rays and their relative coefficients were evaluated and presented. Measured results were compared with those previously measured for other concretes. The results show that ilmenite–limonite concrete is a better reactor biological shield than the other three concretes. Serpentine concrete under investigation is a better reactor fast neutrons shield than ordinary and hematite–serpentine concretes. Serpentine concrete

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

Gamma dose from activation of internal shields in IRIS reactor.

Science.gov (United States)

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

Gamma dose from activation of internal shields in IRIS reactor

International Nuclear Information System (INIS)

Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

Reactor vessel head permanent shield

International Nuclear Information System (INIS)

Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

1989-01-01

A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

Thermal shielding device in LMFBR type reactors

International Nuclear Information System (INIS)

Nakamura, Hiroshi.

1985-01-01

Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)

Technical specifications for the bulk shielding reactor

International Nuclear Information System (INIS)

1986-05-01

This report provides information concerning the technical specifications for the Bulk Shielding Reactor. Areas covered include: safety limits and limiting safety settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and monitoring of airborne effluents. 10 refs

Photon spectrum behind biological shielding of the LVR-15 research reactor

Energy Technology Data Exchange (ETDEWEB)

Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

2011-07-01

The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

Radiological characterization of the concrete biological shield of the APSARA reactor

OpenAIRE

Srinivasan Priya; Srinivasan Panchapakesan; Thomas Shibu; Gopalakrishnan R.K.; Goswami A.

2013-01-01

The first Indian research reactor, APSARA, was utilized for various R&D programmes from 1956 until its shutdown in 2009. The biological shield of the reactor developed residual activity due to neutron irradiation during the operation of the reactor. Dose rate mapping and in-situ gamma spectrometry of the concrete structures of the reactor pool were carried out. Representative concrete samples collected from various locations were subjected to high-resolution gamma spectrometry analysis....

Neutronic reactor thermal shield

International Nuclear Information System (INIS)

Lowe, P.E.

1976-01-01

A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

Core test reactor shield cooling system analysis

International Nuclear Information System (INIS)

Larson, E.M.; Elliott, R.D.

1971-01-01

System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

International Nuclear Information System (INIS)

Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

1982-01-01

During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

Technical specifications: Tower Shielding Reactor II

International Nuclear Information System (INIS)

1979-02-01

The technical specifications define the key limitations that must be observed for safe operation of the Tower Shielding Reactor II (TSR-II) and an envelope of operation within which there is reasonable assurance that these limits cannot be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

Shielding assessment of the ETRR-1 Reactor Under power upgrading

Energy Technology Data Exchange (ETDEWEB)

Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1997-12-31

The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

Nuclear data requirements for fusion reactor shielding

International Nuclear Information System (INIS)

Abdou, M.A.

1979-01-01

The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

Status of reactor shielding research in the United States

International Nuclear Information System (INIS)

Bartine, D.E.

1983-01-01

Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Alternative methodology for irradiation reactor experimental shielding calculation

International Nuclear Information System (INIS)

Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

1996-01-01

Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

International Nuclear Information System (INIS)

Meuschke, R.E.; Boyd, C.H.

1989-01-01

This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

International Nuclear Information System (INIS)

Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

1990-01-01

1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

Tools and applications for core design and shielding in fast reactors

International Nuclear Information System (INIS)

Rachamin, Reuven

2013-01-01

Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

Neutron activation measurements in research reactor concrete shield

International Nuclear Information System (INIS)

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

2001-01-01

The results of activation measurement inside TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete together with gold and nickel foils were irradiated in the reactor body. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active longlived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale.(author)

FFTF reactor-characterization program: gamma-ray measurements and shield characterization

International Nuclear Information System (INIS)

Bunch, W.L.; Moore, F.S. Jr.

1983-02-01

A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA)

Uranium self-shielding in fast reactor blankets

Energy Technology Data Exchange (ETDEWEB)

Kadiroglu, O.K.; Driscoll, M.J.

1976-03-01

The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

Seismic and cask drop excitation evaluation of the tower shielding reactor

International Nuclear Information System (INIS)

Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N.

1989-01-01

During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs

Seismic and cask drop excitation evaluation of the Tower Shielding Reactor

International Nuclear Information System (INIS)

Stover, R.L.; Harris, S.P.; Johnson, J.J.; Sumodobila, B.N.

1989-01-01

During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations

Improvement of top shield analysis technology for CANDU 6 reactor

International Nuclear Information System (INIS)

Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

1996-07-01

As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

Computational methodology for the Oak Ridge Research Reactor (ORR) and Bulk Shielding Reactor (BSR): cross-section and validation. Volume 1

International Nuclear Information System (INIS)

Miller, L.F.; Williams, M.L.

1986-03-01

A neutronics library suitable for low-enrichment uranium (LEU) and high-enrichment uranium (HEU) fueled cores for both the Oak Ridge Research Reactor (ORR) and the Bulk Shielding Reactor (BSR) is documented herein. The library is obtained from version V of the Evaluated Nuclear Data File (ENDF/B-V) and contains 223 nuclides weighted over a variety of region-dependent neutron spectra. Self-shielding and zone-weighting effects are incorporated with 227-group calculations for several reactor-core configurations. Libraries are archived for both transport and diffusion theory seven-group calculations. Complete listings of processing details are included so that libraries with different specifications can be easily obtained. Results from validation calculations indicate that the neutronics libraries obtained from this effort are suitable for neutronics computations for the ORR and BSR. 12 refs., 5 figs., 15 tabs

Radiation shielding provided by residential houses in Japan in reactor accidents accompanied with atmospheric release

International Nuclear Information System (INIS)

Yamaguchi, Yasuhiro; Minami, Kentaro

1991-01-01

The present report describes the radiation shielding effect of houses in Japan against the radioactive cloud resulting from a major reactor accident accompanied with atmospheric release. The shielding factor of houses, the ratio of indoor exposure rate to outdoor one, has been studied for the semi-infinite and finite clouds which contain γ-emitting radionuclides released from a reactor facility. The shielding factor of houses against γ-rays from the radioactive cloud decreases gradually with release delay time and keeps a minimum during the period from 50 to 1000 hours after reactor shutdown while 133 Xe predominates in the cloud. Radioiodines mixed in the cloud raise slightly the shielding factor, and the factor depends little on the shape of the cloud. A set of shielding factors for the use of emergency planning was consequently proposed as 0.4 for simple ferroconcrete residential house and 0.9 for other ordinary ones. (author)

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

International Nuclear Information System (INIS)

Reid, Robert S.; Pearson, J. Bosie; Stewart, Eric T.

2007-01-01

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

International Nuclear Information System (INIS)

Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

2007-01-01

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

Energy Technology Data Exchange (ETDEWEB)

REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

2007-01-16

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Neutron and gamma-ray spectra measurement on the model of the KS-150 reactor radial shielding

International Nuclear Information System (INIS)

Holman, M.; Hogel, J.; Marik, J.; Kovarik, K.; Franc, L.; Vespalec, R.

1977-01-01

A shortened model of the peripheral region of the KS-150 reactor core consisting of two rows of fuel elements and a reflector was constructed from the peripheral fuel elements of the KS-150 reactor core in an experiment on the TR-0 reactor. The mockup of the thermal shield (10 cm of steel), the pressure vessel (15 cm of steel) and the inner wall of the water biological shielding (2 cm of steel) of the KS-150 reactor were erected outside the TR-0 vessel. Fast neutron and gamma spectra were measured with a stilbene crystal scintillation spectrometer. The resonance neutron spectra were measured with 197 Au, 63 Cu and 23 Na resonance activation detectors. Fast neutron spectra inside the reactor were measured with a 10 mm diameter by 10 mm thick stilbene crystal spectrometer, outside the reactor with a 10 mm diameter by 10 mm thick and a 20 mm diameter by 20 mm thick stilbene crystal spectrometer. Neutron spectra in the energy regions of 1 eV to 3 keV and 0.6 MeV to 0.8 MeV were obtained on the core periphery, on the reflector half-thickness and in front of and behind the reactor thermal shield. Gamma spectra were obtained in front of and behind the thermal shield. It was found that the attenuation of neutron fluxes by the reflector and the thermal shield increased with increasing energy while gamma radiation attenuation decreased with increasing energy. It was not possible to obtain the neutron spectrum in the 10 to 600 keV energy range because suitable detection instrumentation was not available. (J.P.)

Fast reactor shield sensitivity studies for steel--sodium--iron systems

International Nuclear Information System (INIS)

Oblow, E.M.; Weisbin, C.R.

1977-01-01

A study was made of the adequacy of the current ENDF/B-IV sodium and iron neutron cross section data files for fast reactor shield design work. Experimental data from 21 fast reactor shield configurations containing large thicknesses of steel, sodium, and iron were analyzed with discrete ordinates calculations and sensitivity methods to assess the data files. This study represents the largest full-scale sensitivity analysis of benchmark quality experimental data to date. Included in the sensitivity studies were the results of the new cross section adjustment algorithms added to the FORSS code system. Conclusions were drawn about the need for more accurate data for sodium and iron elastic and discrete inelastic cross sections above 1 MeV and the values of the total cross section in the vicinity of important minima

FBR type reactors

International Nuclear Information System (INIS)

Nakamura, Tsugio.

1986-01-01

Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

A Shielding Analysis of Hot Cell for a 10 MW Research Reactor

International Nuclear Information System (INIS)

Alnajjar, Alaaddin; Park, Chang Je; Roh, Gyuhong; Lee, Byunchul

2013-01-01

In this paper, a shielding analysis has been performed for the hot cell in a 10 MW research reactor. Two kinds of shielding analysis code systems are used such as MCNPX2.7 and M-Shield7. The first one is Monte Carlo stochastic code and the second one is a deterministic point kernel code. The results are compared in this study. In order to obtain source term, the ORIGEN-S code is used for different kinds of source. Four kinds of sources are taken into consideration. From the simulation, it is also proposed that the proper thickness of shielding material and the maximum source capacity in the hot cell. This study shows preliminary analysis results of hot cell shielding for 10MW research reactor. Total four different source terms are considered such as spent fuel assembly, Ir-192, Mo-99, and I-131. For shielding material, general concrete, heavy concrete, and lead are used. MCNPX code is mainly used for a simplified hot cell model and the result are nearly consistent when compared with M-Shield code. Required shielding thickness and the hot cell capacity are also obtained for various criterion of surface dose rates

Upper shielding body in LMFBR type reactors

International Nuclear Information System (INIS)

Shoji, Koichi.

1986-01-01

Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

Basic nuclear data and reactor shielding design formulaire PROPANE Do

International Nuclear Information System (INIS)

Estiot, J.C.; Salvatores, M.; Trapp, J.P.

1979-01-01

This paper presents a calculational scheme - formulaire PROPANE - to calculate the deep neutron penetration in the fast reactor shield. The emphasis is put on the multigroup data and method approximations. The performances of this formulaire are presented

Shield materials recommended for space power nuclear reactors

Science.gov (United States)

Kaszubinski, L. J.

1973-01-01

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

International Nuclear Information System (INIS)

Seki, Y.; Mori, S.

1984-01-01

The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

Shielding efficiency of metal hydrides and borohydrides in fusion reactors

DEFF Research Database (Denmark)

Singh, Vishvanath P.; Badiger, Nagappa M.; Gerward, Leif

2016-01-01

at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable...... combination of low-and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors....

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

International Nuclear Information System (INIS)

Guthrie, J.A.S.

1980-04-01

In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

The computer code system for reactor radiation shielding in design of nuclear power plant

International Nuclear Information System (INIS)

Li Chunhuai; Fu Shouxin; Liu Guilian

1995-01-01

The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

Nuclear reactor cavity streaming shield

International Nuclear Information System (INIS)

Klotz, R.J.; Stephen, D.W.

1978-01-01

The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

Nuclear reactor assembly

International Nuclear Information System (INIS)

Dorner, H.; Scholz, M.; Jungmann, A.

1975-01-01

A nuclear reactor assembly includes a reactor pressure tank having a substantially cylindrical side wall surrounded by the wall of a cylindrical cavity formed by a biological shield. A rotative cylindrical wall is interposed between the walls and has means for rotating it from outside of the shield, and a probe is carried by the rotative wall for monitoring the pressure tank's wall. The probe is vertically movable relative to the rotative cylindrical wall, so that by the probe's vertical movement and rotation of the rotative cylinder, the reactor's wall can be very extensively monitored. If the reactor pressure tank's wall fails, it is contained by the rotative wall which is backed-up by the shield cavity wall. (Official Gazette)

Nuclear reactor

International Nuclear Information System (INIS)

Scholz, M.

1976-01-01

An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

Activation of concrete samples from the biological shield of the ASTRA reactor

International Nuclear Information System (INIS)

Smecka, F.

2006-09-01

Drill cores from the biological shield of the ASTRA reactor in Seibersdorf were taken and milled because of the different size of the Baryt crystals in the concrete in order to get homogenous samples. The powder samples were put into bore holes of a graphite block which was placed into the thermal column of the TRIGA Mark II reactor. The block was irradiated for 10 minutes at a reactor power of 25 kW. After one hour the dose rate was examined and the samples were ready for further save handling. The gamma spectrum was measured with a Ge detector and the results were compared with simulation data. (nevyjel)

Methodology of shielding calculation for nuclear reactors

International Nuclear Information System (INIS)

Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

1982-01-01

A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

International Nuclear Information System (INIS)

Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2014-01-01

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Concept of spatial channel theory applied to reactor shielding analysis

International Nuclear Information System (INIS)

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

1977-01-01

The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response. The method is analogous to the channel-theory method developed for ascertaining important energy channels in cross-section analysis. The mathematical basis for the theory is shown to be the generalized reciprocity relation, and sample problems are given to exhibit and verify properties predicted by the mathematical equations. A practical example is cited from the shielding analysis of the Fast Flux Test Facility performed at Oak Ridge National Laboratory, in which a perspective plot of channel-theory results was found useful in locating streaming paths around the reactor cavity shield

Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

International Nuclear Information System (INIS)

Merz, Stefan; Djuricic, Mile; Villa, Mario; Boeck, Helmuth; Steinhauser, Georg

2011-01-01

The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10 9 cm -2 s -1 at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: → Neutron activation is an important process for the waste management of nuclear facilities. → Biological shield of the TRIGA reactor Vienna has been topic of investigation. → Flux values allow a categorization of the concrete concerning radiation protection legislation. → Reactor installations are of great importance as neutron sources into the biological shield. → Every installation shows distinguishable flux profiles.

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Resonance shielding in thermal reactor lattices

International Nuclear Information System (INIS)

Rothenstein, W.; Taviv, E.; Aminpour, M.

1982-01-01

The theoretical foundations of a new methodology for the accurate treatment of resonance absorption in thermal reactor lattice analysis are presented. This methodology is based on the solution of the point-energy transport equation in its integral or integro-differential form for a heterogeneous lattice using detailed resonance cross-section profiles. The methodology is applied to LWR benchmark analysis, with emphasis on temperature dependence of resonance absorption during fuel depletion, spatial and mutual self-shielding, integral parameter analysis and treatment of cluster geometry. The capabilities of the OZMA code, which implements the new methodology are discussed. These capabilities provide a means against which simpler and more rapid resonance absorption algorithms can be checked. (author)

Design of a management information system for the Shielding Experimental Reactor ageing management

International Nuclear Information System (INIS)

He Jie; Xu Xianhong

2010-01-01

The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

Design of a management information system for the Shielding Experimental Reactor ageing management

Energy Technology Data Exchange (ETDEWEB)

He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

2010-01-15

The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

Improved Nuclear Reactor and Shield Mass Model for Space Applications

Science.gov (United States)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

Directory of Open Access Journals (Sweden)

GO CHIBA

2014-06-01

Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Study of filtration of reactor beam of neutrons with cadmium in a multilayer shielding containing boron carbide

International Nuclear Information System (INIS)

Megahid, R.M.; El-Kall, E.H.

1986-01-01

Experimental measurements were carried out to study the effect of cadmium on the distribution and attenuation of reactor thermal neutrons emitted from a reactor core and the new thermal neutrons produced in a heterogeneous shield of water, iron, iron + B 4 C and ordinary concrete. The measurements were made using a reactor beam of neutrons filtered with cadmium emitted from one of the horizontal channels of ET-RR-1. It is found that the presence of cadmium sheet at channel exit causes a marked decrease in the thickness of the shield required to attenuate the thermal neutron flux by a certain factor. 12 refs., 5 figures. (author)

Method of shielding a liquid-metal-cooled reactor

International Nuclear Information System (INIS)

Sayre, R.K.

1978-01-01

The primary heat transport system of a nuclear reactor - particularly for a liquid-metal-cooled fast-breeder reactor - is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of a the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system

Application of MCNP code in shielding calculation of minitype fast reactor

International Nuclear Information System (INIS)

He Keyu; Han Weishi

2008-01-01

An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)

Neutronics shielding analysis of the last mirror-beam duct system for a laser fusion power reactor

International Nuclear Information System (INIS)

Ragheb, M.M.H.; Klein, A.C.

1981-01-01

A Monte Carlo three-dimensional neutronics analysis for the last mirror-beam duct system for the SOLASE conceptual laser-driven fusion power reactor design is presented. Detailed geometric configurations including the reactor cavity, the two last mirrors, and the three-section two-right-angle bends duct are modeled. Measurements are given of the dimensions and compositions of the reactor components, and of neutron scalar fluxes, spatial dependencies and neutron volumetric heating rates for the cases of aluminum or Boral as laser beam duct liners, and ordinary concrete or lead mortar as shield material. A three-dimensional modeling of laser-driven reactor penetrations is employed. The particle leakage is found to be excessively high for the configuration of the conceptual design considered and the advantages and disadvantages of various solutions, such as the use of Boral as a duct liner and the use of lead mortar instead of ordinary concrete as a shield material, are considered

Assessment of the implementation of a neutron measurement system during the commissioning of the Jordan Research and Training Reactor

Energy Technology Data Exchange (ETDEWEB)

Bae, Sang Hoon; Suh, Sang Mun [Division of Research Reactor System Design, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cha, Han Ju [Dept. of Electrical Engineering, Intelligent Power Conversion Laboratory, Chungnam National University, Daejeon (Korea, Republic of)

2017-04-15

The Jordan Research and Training Reactor (JRTR) is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS) applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

Structure shielding from cloud and fallout gamma ray sources for assessing the consequences of reactor accidents

International Nuclear Information System (INIS)

Burson, Z.G.; Profio, A.E.

1975-12-01

Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents

Embrittlement of the Shippingport reactor shield tank

International Nuclear Information System (INIS)

Chopra, O.K.; Shack, W.J.

1989-01-01

Surveillance specimens from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory showed an unexpectedly high degree of embrittlement relative to the data obtained on similar materials in Materials Testing Reactors (MTRs). The results suggest a possible negative flux effect and raise the issue of embrittlement of the pressure vessel support structures of commercial light water reactors. To help resolve this issues, a program was initiated to characterize the irradiation embrittlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport NST operated at 55 degree C (130 degree F) and was fabricated from A212 Grade B steel, similar to the vessel material in HFIR. The inner wall of the NST was exposed to a total maximum fluence of ∼ 6 x 10 17 n/cm 2 (E > 1 MeV) over a life of 9.25 effective full power years. This corresponds to a fast flux of 2.1 x 10 9 n/cm 2 x s and is comparable to the conditions for the HFIR surveillance specimens. The results indicate that irradiation increases the 15 ft x lb Charpy transition temperature (CTT) by ∼25 degree C (45 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens and is consistent with that expected from the MTR data base. However, the actual value of CTT is high, and the toughness at service temperature is low, even when compared with the HFIR data. The increase in yield stress is ∼50 MPa, which is comparable to the HFIR data. The results also indicate a lower impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall

Apparatus for sealing a rotatable shield plug in a liquid metal nuclear reactor

International Nuclear Information System (INIS)

Winkleblack, R.K.

1980-01-01

An apparatus for sealing a rotatable shield plug in a nuclear reactor having liquid metal coolant is described. The apparatus includes a dip -ring seal adapted to provide a fluid barrier between the liquid metal and the atmosphere and to permit rotation of the shield plug. The apparatus also includes a static seal for the rotatable shield plug located between the dip-ring seal and the liquid metal. The static seal isolates the dip-ring seal from the liquid metal vapor during operation at power and can be disengaged for rotation of the shield plug

Nuclear reactor installation

International Nuclear Information System (INIS)

Keller, W.

1976-01-01

A nuclear reactor installation includes a pressurized-water coolant reactor vessel and a concrete biological shield surrounding this vessel. The shield forms a space between it and the vessel large enough to permit rapid escape of the pressurized-water coolant therefrom in the event the vessel ruptures. Struts extend radially between the vessel and shield for a distance permitting normal radial thermal movement of the vessel, while containing the vessel in the event it ruptures, the struts being interspaced from each other to permit rapid escape of the pressurized-water coolant from the space between the shield and the vessel

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

Analysis of crack-formation in the shielding concrete of a TRIGA Mark II reactor

International Nuclear Information System (INIS)

Linsbauer, H.; Maydl, P.

1978-01-01

Within a short time after the start-up of the reactor several cracks appeared at the concrete surface and the number and width of the cracks had grown till now. Experimental and theoretical analysis were made in order to investigate the origin of the cracks and to prevent further crack increase. Crack movement was measured by inductive gages and simultaneously the temperature of the cooling water in the reactor tank at the top and at the bottom as well as the air and the concrete temperature were recorded. The calculations of the thermal stresses were made in two independent ways: 1. Analytically, simulating the shielding concrete as an infinite hollow cylinder of constant thickness and 2. Using the Finite Element method, for a better description of the geometry. It was concluded that the cracks of the shielding concrete are exclusively caused by the thermal stresses. The thermal insulation at the lower part of the shielding is not effective. The structural system of the shielding concrete as a monolithic block without joints produces automatically tensile stresses

ITER [International Thermonuclear Experimental Reactor] reactor building design study

International Nuclear Information System (INIS)

Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

Multi-objective optimization of a compact pressurized water nuclear reactor computational model for biological shielding design using innovative materials

Energy Technology Data Exchange (ETDEWEB)

Tunes, M.A., E-mail: matheus.tunes@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil); Oliveira, C.R.E. de, E-mail: cassiano@unm.edu [Department of Nuclear Engineering, The University of New Mexico, Farris Engineering Center, 221, Albuquerque, NM 87131-1070 (United States); Schön, C.G., E-mail: schoen@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil)

2017-03-15

Highlights: • Use of two n-γ transport codes leads to optimized model of compact nuclear reactor. • It was possible to safely reduce both weight and volume of the biological shielding. • Best configuration obtained by using new composites for both γ and n attenuation. - Abstract: The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

International Nuclear Information System (INIS)

Toshimas, M.; Nobuo, S.

1983-01-01

For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

Waste management for JAERI fusion reactors

International Nuclear Information System (INIS)

Tobita, K.; Nishio, S.; Konishi, S.; Jitsukawa, S.

2004-01-01

In the fusion reactor design study at Japan Atomic Energy Institute (JAERI), several waste management strategies were assessed. The assessed strategies are: (1) reinforced neutron shield to clear the massive ex-shielding components from regulatory control; (2) low aspect ratio tokamak to reduce the total waste; (3) reuse of liquid metal breeding material and neutron shield. Combining these strategies, the weight of disposal waste from a low aspect ratio reactor VECTOR is expected to be comparable with the metal radwaste from a light water reactor (∼4000 t)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

Methods of preventing fast breeder reactor shield plug from adhesion of sodium

International Nuclear Information System (INIS)

Hashiguchi, Koh; Hara, Johji; Nei, Hiromichi; Daiku, Motoichi; Wagatsuma, Kenji

1980-01-01

The shield plug, which is located at the upper part of a reactor vessel of a sodium-cooled fast breeder reactor, is composed of a rotating and a stationary plug. Fuel exchange is performed easily by the rotation of the rotating plug. The vapor or mist of sodium evaporated from liquid sodium deposits on the gap surfaces of the rotating and stationary plugs and is solidified or changed into a solid reactant. If such condition continues for a long period, harmful effects are exerted on the fuel exchange operation. In order to develop methods of preventing the sodium deposition, investigation was made on the phenomenon of sodium deposition. By the use of the testing equipment simulating the shield plug, deposition tests and specimen measurements were made for different gap width test section size and condition. On the basis of the effects of these parameters clarified by experiments, the effectiveness of three kinds of mechanism for preventing sodium deposition were investigated experimentally. In addition, by using a thermo-siphon analogical model, analysis was performed to deduce experimental equations for sodium deposition. (author)

Characterisation of the inventory of radioisotopes induced in the biological shield a WWER-440 reactor

International Nuclear Information System (INIS)

Feher, S.; Czifrus, Sz.; Zsolnay, E.M.; Szondi, E.

2001-01-01

A significant part of the radwaste originating from the decommissioning of NPPs is made up of the activated concrete and steel components of the biological shield. The paper presents the results of studies aimed at the determination of the amount of radionuclides accumulating in the serpentinous and ordinary concrete shield around the WWER-440 reactors of the Paks NPP. For the calculations, the reactor, vessel and shield were modelled in detail both in terms of geometry and material composition. The spatial and energy distribution of the activating neutron spectrum was determined by certain modules of SCALE 4.3 and the code TORT in two and three dimensions, while the activation was calculated using ORIGEN-S for 22 geometrical regions. The results showed that the activity of the concrete structures at final shutdown after 30 years of operation is approximately 50 TBq, which decreases to 20, 12, 1.1 TBq and 27 GBq after 1 month, 1 year, 10 and 100 years, respectively (Authors)

Heat-pipe thermionic reactor concept

DEFF Research Database (Denmark)

Storm Pedersen, E.

1967-01-01

Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

1977-09-01

Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1984-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1976-09-01

Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1980-09-01

Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Fusion reactor design studies

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

1990-01-01

This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1975-11-01

Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

International Nuclear Information System (INIS)

Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

2012-01-01

Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

Dose rate in the reactor room and environment during maintenance in fusion reactors

International Nuclear Information System (INIS)

Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

1995-01-01

According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

International Nuclear Information System (INIS)

Souza, Gregorio Soares de

2011-01-01

The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10 B (n, α) 7 Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

Integrated thermal analysis of top-shield and reactor vault of Indian FBR-600

International Nuclear Information System (INIS)

Rajendrakumar, M.; Velusamy, K.; Selvaraj, P.

2015-01-01

The design for next generation fast breeder reactors (FBR-600) has been commenced with enhanced safety and improved economy as the main targets. The Top Shield (TS) of Prototype Fast Breeder Reactor (PFBR) is a box type structure consisting of Roof Slab (RS), Small Rotatable Plug (SRP), and Large Rotatable Plug (LRP). The large box type structure with many penetrations posed difficulties during manufacturing. Because of the required high load carrying capabilities, a dome shaped thick plate roof slab is conceived for FBR-600. Main Vessel (MV) which holds the primary sodium and associated components is welded to the RS through a triple joint. Reactor vault (RV) is a thick concrete structure which supports MV and Safety Vessel (SV). The temperature of RV concrete has to be less than 338 K (65°C) under normal operating heat loads (full and part load conditions) and less than 363 K (90°C) under Safety Grade Decay Heat Removal (SGDHR) conditions with one cooling loop in service. The temperature in the component penetrations of the RS should be greater than 120°C to avoid sodium aerosol deposition. Similarly, the temperature of the LRP and SRP has to be ∼120°C to protect the elastomeric seals provided to these structures. Further, the heat load to RV transferred by direct conduction by roof slab support has to be minimum. To meet these conflicting thermal requirements, detailed multi-physics CFD calculations have been performed to finalize, (i) the insulation requirements on the top of roof slab, (ii) number and position of reflective insulation plates below the bottom plate of roof slab/rotating plugs, (iii) air flow rate for various zones of the top shield and (iv) water flow rate and pitch of water cooling pipes for the reactor vault. (author)

TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

International Nuclear Information System (INIS)

Lee, J.D.; Sawan, M.

1987-01-01

TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

Experience with reactor assembly of FBTR

International Nuclear Information System (INIS)

Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

2006-01-01

Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

Directory of Open Access Journals (Sweden)

Ersez Tunay

2017-01-01

Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

Science.gov (United States)

Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

2017-09-01

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Comparison of calculational methods for liquid metal reactor shields

International Nuclear Information System (INIS)

Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

1985-09-01

A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

SP-100 reactor cell activation

International Nuclear Information System (INIS)

Wilcox, A.D.

1991-09-01

There are plans to test the SP-100 space reactor for 2 yr in the test facility shown in Figure 1. The vacuum vessel will be in the reactor experiment (RX) cell surrounded by an inert gas atmosphere. It is proposed that the reactor test cell could contain removable-water- shielding tanks to reduce the residual activation dose rates in the test cell after the tests are completed. This reduction will allow the facility to be considered for other uses after the SP-100 tests are completed. The radiation dose rates in the test cell were calculated for several configurations of water-shielding tanks to help evaluate this concept

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

Nuclear characteristics of D-D fusion reactor blankets

International Nuclear Information System (INIS)

Nakashima, Hideki; Ohta, Masao

1978-01-01

Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

Upgradation of Apsara reactor

International Nuclear Information System (INIS)

Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

2009-01-01

Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

Nuclear design of a very-low-activation fusion reactor

International Nuclear Information System (INIS)

Cheng, E.T.; Hopkins, G.R.

1983-06-01

An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Nuclear reactor

International Nuclear Information System (INIS)

Jungmann, A.

1975-01-01

Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

Matsuura, S.; Nakahara, Y.; Takano, H.

1983-09-01

Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Adjustment equipment for reactor radioactivity meter

International Nuclear Information System (INIS)

Denisov, V.P.; Malishev, A.N.; Shebanova, L.E.; Kirilyuk, N.A.; Maksimov, Yu.N.; Bessalov, G.G.; Vikhorev, Yu.V.; Lukyanov, M.A.

1978-01-01

An activity meter is described movably located in a channel placed in the peripheral biological shielding of a nuclear reactor. It is connected to a weight moving in a second channel by means of a pulley. This arrangement allows locating the radioactivity meter drive on the outer side of the biological shield and vacating space above the reactor body. (Ha)

Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-09-01

The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

Demonstration test on manufacturing 200 l drum inner shielding material for recycling of reactor operating metal scrap

International Nuclear Information System (INIS)

Umemura, A.; Kimura, K.; Ueno, H.

1993-01-01

Low-level reactor wastes should be safely recycled considering those resource values, the reduction of waste disposal volume and environmental effects. The reasonable recycling system of reactor operating metal scrap has been studied and it was concluded that the 200 liter drum inner shielding material is a very promising product for recycling within the nuclear industry. The drum inner shielding material does not require high quality and so it is expected to be easily manufactured by melting and casting from roughly sorted scrap metals. This means that the economical scrap metal recycling system can be achieved by introducing it. Furthermore its use will ensure safety because of being contained in a drum. In order to realize this recycling system with the drum inner shielding material, the demonstration test program is being conducted. The construction of the test facility, which consists of a melting and refining furnace, a casting apparatus, a machining apparatus etc., was finishing in September, 1992

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

International Nuclear Information System (INIS)

Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

2015-01-01

In ICONE22, the authors presented the Integrated Head Assembly (IHA) design concept implemented at Callaway Nuclear Power Plant in Missouri, USA. The IHA concept is implemented to reduce the outage duration and the associated radiation exposure to the workers by reducing critical path time during Plant Refueling Outage. One of the head area components in the IHA is a steel missile shield designed to protect the Control Rod Drive Mechanism (CRDM) assembly from damaging other safety-related components in the vicinity in the Containment. Per Federally implemented General Design Criteria for commercial nuclear plants in the USA, the design of Nuclear Steam Supply System (NSSS) must provide protection from the damages caused by a postulated event of CRDM housing units and their associated parts disengaging from the reactor vessel assembly. This event is considered as a Loss of Coolant Accident (LOCA) and assumes that once the CRDM housing unit and their associated parts disengage from the reactor vessel internals assembly, they travel upward by the water jet with the following sequence of events: Per Reference 1, the drive shaft and control rod cluster are forced out of the reactor core by the differential pressure across the drive shaft with the assumption that the drive shaft and control rod cluster, latched together, are fully inserted when the accident occurs. After the travel, the rod cluster control spider will impact the lower side of the upper support plate inside the reactor vessel fracturing the flexure arms in the joint freeing the drive shaft from the control rod cluster. The control rod cluster is stopped by the upper support plate and will remain below the upper support plate during this accident. However, the drive shaft will continue to accelerate in the upward direction until it is stopped by a safety feature in the IHA. The integral missile shield as a safety feature in the IHA is designed to stop the CRDM drive shaft from moving further up in the

Nuclear design of the blanket/shield system for a Tokamak Experimental Power Reactor

International Nuclear Information System (INIS)

Abdou, M.A.

1976-01-01

The various options and trade-offs in the nuclear design of the blanket/shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, Δ/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on Δ/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest Δ/sub BS/ for a given attenuation. This Δ/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m 2 neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/shield and magnet are well within the practical limits for heat removal

Nuclear data needs for fast breeder reactor shielding

International Nuclear Information System (INIS)

Oblow, E.M.; Perey, F.G.

1978-11-01

A review of neutron and gamma-ray cross section data needs for fast reactor shielding is presented in light of the recent advances made in assessing these needs through sensitivity studies. Total and partial cross sections and energy and angular distribution data for neutrons are surveyed as well as gamma-ray production cross sections. The strengths and deficiencies of currently available benchmark-quality integral experiments are also discussed with respect to their use in creating adjusted cross section libraries for design work. The availability of first round covariance data in ENDF/B-IV and plans for ENDF/B-V are also reviewed. This latter information makes it possible to quantitatively assess the quality of current cross section data libraries and also puts adjustment and data assessment procedures on a firmer basis

Evaluation of some resonance self-shielding procedures employed in high conversion light water reactor design

International Nuclear Information System (INIS)

Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

1990-01-01

The procedures employed in the treatment of the resonance shielding effect have been identified as one of the causes of the large discrepancies found in the neutronic calculation of high conversion light water reactors (HCLWRs), indicating the need for a revision of the self-shielding procedures employed. In this work some well known techniques applied in HCLWR self-shielding calculations are evaluated; the study involves the comparison of methods for the generation of group constants, the analysis of the impact of considering some isotopes as infinitely diluted and the evaluation of the usual approximations utilized for the treatment of heterogeneities

Nuclear reactor building

International Nuclear Information System (INIS)

Oshima, Nobuaki.

1991-01-01

The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

Problems related to the definition of the shielding of a large fast power reactor

International Nuclear Information System (INIS)

Moreau, J.

Solutions for the shielding of a 1000 MW(e) power plant in the same technological line as Phenix are given. They have been evaluated with a monodimensional transport code. The choice is based on the comparison of their efficiency towards neutrons and on the consequences of their characteristics on the conception of the reactor tank. A few economical considerations give an idea of the influence of the choice in shielding on the cost of the power plant. At last the problem of the optimization possibilities is approached from the designer's point of view

Neutronic reactor

International Nuclear Information System (INIS)

Wende, C.W.J.

1976-01-01

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1975-02-01

This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor engineering department annual report

International Nuclear Information System (INIS)

1990-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Lateral restraint assembly in a nuclear reactor

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, W.

1977-01-01

A lateral restraint assembly is described for a reactor of, for example, the high temperature gas-cooled type which commonly includes a reactor core of relatively complex construction supported within a shell or vessel providing a shielded cavity for containing the reactor core. (U.K.)

Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

Energy Technology Data Exchange (ETDEWEB)

Nilsson, R; Aalto, E

1964-09-15

Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

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

Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding)

International Nuclear Information System (INIS)

1978-01-01

An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976

Design of the segment structure and coolant ducts for a fusion reactor blanket and shield

International Nuclear Information System (INIS)

Briaris, D.A.; Stanbridge, J.R.

1978-05-01

An outline design and analysis of a support structure for the replaceable first wall of a helium cooled fusion reactor blanket has been undertaken. The proposed structure supports all the segment gravitational loads with maximum deflections limited to < 10 mm, and is itself supported off the outer shield by a simple vee-in-groove arrangement. It is a feature of the design that the coaxial coolant pipes and the segment structure operate at the same temperature, making it possible for them to be integrated, thereby avoiding the necessity for pipe bellows. The requirements of cooling the inner arm of the structure and increasing the major radius of the torus by approximately = 0.5 m, have been identified as problems associated with the 'horseshoe' shaped structure applicable to the reactor with divertor. For a ring structure, i.e. reactor without divertor, these problems do not arise. (author)

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Tritium monitoring within the reactor hall of a DT fusion reactor

International Nuclear Information System (INIS)

Jalbert, R.A.

1983-01-01

Monitoring the reactor hall atmosphere of DT-fueled fusion reactors will probably be performed with conventional ion chamber and proportional counter instruments modified as necessry to deal with the background radiation. Background includes external neutron and gamma radiation and internal beta-gamma radiation from the activated atmosphere. Although locating instruments in remote areas of the reactor hall and adding local shielding and electronic compensation may be feasible, placing the instruments in accessible low-background areas outside of the reactor hall and doing remote sampling is preferable and solves most of the radiation problems. The remaining problem of the activated atmosphere may be solved by recently developed instruments in conjunction with the use of semi-permeable membranes currently under development and evaluation

Nuclear data needs for fusion reactors

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

International Nuclear Information System (INIS)

Mohd Rafi Mohd Solleh; Abdul Aziz Tajuddin; Abdul Aziz Mohamed; Eid Mahmoud Eid Abdel Munem; Mohamad Hairie Rabir; Julia Abdul Karim; Yoshiaki, Kiyanagi

2011-01-01

The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 10 8 n/ cm 2 / s. According to IAEA (2001) flux of 1.00 x 10 9 n/ cm 2 / s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Activation of TRIGA Mark II research reactor concrete shield

International Nuclear Information System (INIS)

Zagar, Tomaz; Ravnik, Matjaz; Bozic, Matjaz

2002-01-01

To determine neutron activation inside the TRIGA research reactor concrete body a special sample-holder for irradiation inside horizontal channel was developed and tested. In the sample-holder various samples can be irradiated at different concrete shielding depths. In this paper the description of the sample-holder, experiment conditions and results of long-lived activation measurements are given. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active long-lived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale. (author)

Reactor physics computations

International Nuclear Information System (INIS)

Shapiro, A.

1977-01-01

Those reactor-core calculations which provide the effective multiplication factor (or eigenvalue) and the stationary (or fundamental mode) neutron-flux distribution at selected times during the lifetime of the core are considered. The multiplication factor is required to establish the nuclear composition and configuration which satisfy criticality and control requirements. The steady-state flux distribution must be known to calculate reaction rates and power distributions which are needed for the thermal, mechanical and shielding design of the reactor, as well as for evaluating refueling requirements. The calculational methods and techniques used for evaluating the nuclear design information vary with the type of reactor and with the preferences and prejudices of the reactor-physics group responsible for the calculation. Additionally, new methods and techniques are continually being developed and made operational. This results in a rather large conglomeration of methods and computer codes which are available for reactor analysis. The author provides the basic calculational framework and discusses the more prominent techniques which have evolved. (Auth.)

On the research activities in reactor and neutron physics using the first egyptian research reactor

International Nuclear Information System (INIS)

Hassan, A.M.

2000-01-01

A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

A tokamak reactor with servicing capability

International Nuclear Information System (INIS)

Mitchell, J.T.D.; Hollis, A.

1976-01-01

A conceptual design for a Tokamak reactor with practical facilities for the regular replacement of blanket components after the inevitable damage from neutron irradiation, and fatigue is described. This essential facility has been largely ignored in published fusion reactor designs. One exception is the inertially-confined Saturn proposal. Tokamak and other toroidal closed-line systems have very complex geometries and sub-system requirements, which result in blanket servicing being a very difficult problem. In the concept described the magnet shield is divided into two structures - an outer permanent one with access doors and an inner shield, part of and supporting the blanket inside. Servicing access is horizontally between the toroidal magnet coils, after moving some outer poloidal magnet coils. The reactor, reactor hall, workshops and remote-handling facilities are described, and the servicing requirements discussed. The important servicing operation is the remote replacement of radiation damaged blanket and shield - divided in this design into 20 sectors, each weighing 75-100 tons and 11-12 metres high. Analysis of the operation indicates that if one sector can be replaced during a single weekend - i.e. a period of low power demand - then the annual reactor-generator availability allowing as well for the general plant servicing should be >0.9. This level of availability should meet the requirements of generating authorities but the facilities, equipment and workshops necessary may be complex and expensive

Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

International Nuclear Information System (INIS)

Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

2000-01-01

Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

NMS Prototype development final report

International Nuclear Information System (INIS)

Lepetich, J.E.

1993-01-01

Program for development of NMS prototype for LAMPF consisted of 5 tasks: crystal procurement specification, inspection/evaluation of CsI crystals, design/fabrication of crystal housing, design/fabrication of PMT shields, and packaging of crystals in the housing

Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

International Nuclear Information System (INIS)

1980-05-01

An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given

Minimum thickness blanket-shield for fusion reactors

International Nuclear Information System (INIS)

Karni, Y.; Greenspan, E.

1989-01-01

A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Power reactor design trends

International Nuclear Information System (INIS)

Hogan, W.J.

1985-01-01

Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

Inertial fusion reactor designs

International Nuclear Information System (INIS)

Meier, W.

1987-01-01

In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

Science.gov (United States)

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

Damage analysis of TRIGA MARK II Bandung reactor tank material structure

International Nuclear Information System (INIS)

Soedardjo; Sumijanto

2000-01-01

Damage of Triga Mark II Bandung reactor tank material structure has been analyzed. The analysis carried out was based on ultrasonic inspection result in 1996 and the monthly reports of reactor operation by random data during 1988 up to 1995. Ultrasonic test data had shown that thinning processes on south and west region of reactor out side wall at upper part of water level had happened. Reactor operation data had shown the demineralized water should be added monthly to the reactor and bulk shielding water tank. Both reactor and bulk shielding tank are shielded by concrete of Portland type I cement consisting of CaO content about 58-68 %. The analysis result shows that the reaction between CaO and seepage water from bulk shielding wall had taken place and consequently the reactor out sidewall surroundings became alkaline. Based on Pourbaix diagram, the aluminum reactor tank made of aluminum alloy 6061 T6 would be corroded easily at pH equal an greater than 8.6. The passive layer AI 2 O 3 aluminum metal surface would be broken due to water reaction taken place continuously at high pH and produces hydrogen gas. The light hydrogen gas would expand the concrete cement and its expanding power would open the passive layer of aluminum metal upper tank. The water sea pages from adding water into reactor tank could indicate the upper water level tank corrosion is worse than the lower water level tank. (author)

Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

1999-09-01

A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

International Nuclear Information System (INIS)

Ward, D.R.; Holland, L.B.

1979-09-01

The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor

International Nuclear Information System (INIS)

Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

2000-01-01

Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding). [1973--1976

Energy Technology Data Exchange (ETDEWEB)

1978-01-01

An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976.

Research reactor decommissioning experience - concrete removal and disposal -

International Nuclear Information System (INIS)

Manning, Mark R.; Gardner, Frederick W.

1990-01-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

Application of SCALE 6.1 MAVRIC Sequence for Activation Calculation in Reactor Primary Shield Concrete

International Nuclear Information System (INIS)

Kim, Yong IL

2014-01-01

Activation calculation requires flux information at desired location and reaction cross sections for the constituent elements to obtain production rate of activation products. Generally it is not an easy task to obtain fluxes or reaction rates with low uncertainties in a reasonable time for deep penetration problems by using standard Monte Carlo methods. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence in SCALE 6.1 code package is intended to perform radiation transport on problems that are too challenging for standard, unbiased Monte Carlo methods. And the SCALE code system provides plenty of ENDF reaction types enough to consider almost all activation reactions in the nuclear reactor materials. To evaluate the activation of the important isotopes in primary shield, SCALE 6.1 MAVRIC sequence has been utilized for the KSNP reactor model and the calculated results are compared to the isotopic activity concentration of related standard. Related to the planning for decommission, the activation products in concrete primary shield such as Fe-55, Co-60, Ba-133, Eu-152, and Eu-154 are identified as important elements according to the comparisons with related standard for exemption. In this study, reference data are used for the concrete compositions in the activation calculation to see the applicability of MAVRIC code to the evaluation of activation inventory in the concrete primary shield. The composition data of trace elements as shown in Table 1 are obtained from various US power plant sites and accordingly they have large variations in quantity due to the characteristics of concrete composition. In practical estimation of activation radioactivity for a specific plant related to decommissioning, rigorous chemical analysis of concrete samples of the plant would first have to be performed to get exact information for compositions of concrete. Considering the capability of solving deep penetration transport problems and richness

Solid-Core, Gas-Cooled Reactor for Space and Surface Power

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Oscillating liquid flow ICF Reactor

International Nuclear Information System (INIS)

Petzoldt, R.W.

1990-01-01

Oscillating liquid flow in a falling molten salt inertial confinement fusion reactor is predicted to rapidly clear driver beam paths of residual liquid droplets. Oscillating flow will also provide adequate neutron and x-ray protection for the reactor structure with a short (2-m) fall distance permitting an 8 Hz repetition rate. A reactor chamber configuration is presented with specific features to clear the entire heavy-ion beam path of splashed molten salt. The structural components, including the structure between beam ports, are shielded. 3 refs., 12 figs

Radiation streaming in power reactors. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

1979-02-01

Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

Conception of thermonuclear reactor with a shielding layer of the first wall

International Nuclear Information System (INIS)

Marin, S.V.

1979-01-01

Considered is the way of the shielding of the first wall of a thermonuclear reactor by the layer of ISSEC (Internal spectral shifter and Energy Converter). It is a constructive non-power element placed between a plasma and the first wall, and intended for the softening of the spectrum and intensity reduction of particle fluxes falling on the first wall. Results of neutron-physical calculations of the UWMAK-type reactor blanket (in the S 4 -P 3 approximation) are presented. While comparing five materials (C, Mo, Nb, V,W) by the rate of radiation damage formation, gas production, radioactivity level and energy output in the blanket with the 316 stainless steel first wall, it is obvious that the conception of ISSEC permits to prolong the service period of the first wall. Construction elements should be then in the same irradiation conditions as those in fast reactors. Molybdenum has been taken as the best ISSEC material. It reduces the number of displaced atoms of the first wall by 20% and decreases helium production by about 100%, increases energy output in the blanket by 15-18%. However, graphite is advantageous, while comparing it to molybdenum in values of residual energy output, radioactivity level, costs and manufacture simplicity. One problem stays unsolved, which is connected with chemical sputtering of graphite at the formation of C 2 H 2 in the high temperature range. So it is hard to prefer any material now

Importance of self-shielding for improving sensitivity coefficients in light water nuclear reactors

International Nuclear Information System (INIS)

Foad, Basma; Takeda, Toshikazu

2014-01-01

Highlights: • A new method has been developed for calculating sensitivity coefficients. • This method is based on the use of infinite dilution cross-sections instead of effective cross-sections. • The change of self-shielding factor due to cross-section perturbation has been considered. • SRAC and SAINT codes are used for calculating improved sensitivities, while MCNP code has been used for verification. - Abstract: In order to perform sensitivity analyzes in light water reactors where self-shielding effect becomes important, a new method has been developed for calculating sensitivity coefficient of core characteristics relative to the infinite dilution cross-sections instead of the effective cross-sections. This method considers the change of the self-shielding factor due to cross-section perturbation for different nuclides and reactions. SRAC and SAINT codes are used to calculate the improved sensitivity; while the accuracy of the present method has been verified by MCNP code and good agreement has been found

Burst shield for a pressurized nuclear-reactor core and method of operating same

International Nuclear Information System (INIS)

Beine, B.; Schilling, F.

1976-01-01

A pressurized nuclear-reactor core stands on a base up from which extends a cylindrical side wall formed of a plurality of hollow iron castings held together by circumferential and longitudinal prestressed elements. A cylindrical space between this shield and the core serves for inspection of the core and is normally filled with cast-iron segmental slabs so that if the core bursts pieces thrown out do not acquire any dangerous kinetic energy before engaging the burst shield. The top of the shield is removably secured to the side so that it can be moved out of the way periodically for removal of the filler slabs and inspection of the core. An anchor on the upper end of each longitudinal prestressing element bears against a sleeve pressing against the uppermost side element, and a nut engageable with this anchor is engageable down over the top to hold it in place, removal of this nut leaving the element prestressed in the side wall. 11 claims, 16 drawing figures

Radiation environment of fusion experimental reactor

International Nuclear Information System (INIS)

Mori, Seiji; Seki, Yasushi

1988-01-01

Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

Reactor vessel dismantling at the high flux materials testing reactor Petten

International Nuclear Information System (INIS)

Tas, A.; Teunissen, G.

1986-01-01

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

Economic analysis of EBT reactor

International Nuclear Information System (INIS)

Woo, J.T.; Uckan, N.A.; Lidsky, L.M.

1977-01-01

In order to establish the economic potential of the Elmo Bumpy Torus (EBT) reactor, two independent system-costing models have been developed. Both models predict capital costs of approximately $400/kW(th). These relatively low costs reflect the simplicity of the EBTR design. In particular, the modular nature of the individual blanket-shield segments, the low costs ''accelerator style'' containment building, high beta, and steady-state operation lead to relatively low reactor costs. A detailed cost breakdown for subsystems is analyzed. High cost and high uncertainty subsystems are identified to direct further emphasis into those areas. The calculated capital costs for the EBT reactor are compared with those costs quoted for tokamak reactors

STARFIRE remote maintenance and reactor facility concept

International Nuclear Information System (INIS)

Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

1981-01-01

A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

NEUTRONIC REACTOR

Science.gov (United States)

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Production of a datolite-based heavy concrete for shielding nuclear reactors and megavoltage radiotherapy rooms

International Nuclear Information System (INIS)

Mortazavi, S. M. J.; Mosleh-Shirazi, M.A.; Baradaran-Ghahfarokhi, M.; Siavashpour, Z.; Farshadi, A.; Ghafoori, M.; Shahvar, A.

2010-01-01

Biological shielding of nuclear reactors has always been a great concern and decreasing the complexity and expense of these installations is of great interest. In this study, we used datolite and galena minerals for production of a high performance heavy concrete. Materials and Methods: Datolite and galena minerals which can be found in many parts of Iran were used in the concrete mix design. To measure the gamma radiation attenuation of the Datolite and galena concrete samples, they were exposed to both narrow and wide beams of gamma rays emitted from a cobalt-60 radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. To test the compression strengths, both types of concrete mixes (Datolite and galena and ordinary concrete) were investigated. Results: The concrete samples had a density of 4420-4650 kg/m 3 compared to that of ordinary concrete (2300-2500 kg/m 3 ) or barite high density concrete (up to 3500 kg/m 3 ). The measured half value layer thickness of the Datolite and galena concrete samples for cobalt-60 gamma rays was much less than that of ordinary concrete (2.56 cm compared to 6.0 cm). Furthermore, the galena concrete samples had a significantly higher compressive strength as well as 20% more neutron absorption. Conclusion: The Datolite and galena concrete samples showed good shielding/engineering properties in comparison with other reported samples made, using high-density materials other than depleted uranium. It is also more economic than the high-density concretes. Datolite and galena concrete may be a suitable option for shielding nuclear reactors and megavoltage radiotherapy rooms.

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)

Summary of ORNL high-temperature gas-cooled reactor program

International Nuclear Information System (INIS)

Kasten, P.R.

1981-01-01

Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

Activity report of Reactor Physics Division - 1993

International Nuclear Information System (INIS)

Indira, R.

1994-01-01

The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs

Activity report of Reactor Physics Division-1995

International Nuclear Information System (INIS)

Gopalakrishnan, V.

1996-01-01

The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1995 are reported. The activity are arranged under the headings: Nuclear Data Processing and Validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. refs., figs., tabs

Activity report of Reactor Physics Division - 1993

Energy Technology Data Exchange (ETDEWEB)

Indira, R [ed.; Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1994-12-31

The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs.

Operational reactor physics analysis codes (ORPAC)

International Nuclear Information System (INIS)

Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

2007-07-01

For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1979-09-01

Research activities in the Division of Reactor Engineering in fiscal 1978 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committees on Reactor Physics and in Decommissioning of Nuclear Facilities. (author)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-02-01

This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

HYLIFE-II reactor chamber design refinements

International Nuclear Information System (INIS)

House, P.A.

1994-06-01

Mechanical design features of the reactor chamber for the HYLIFE-II inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams (Li 2 BeF 4 ) are used for shielding and blast protection of the chamber walls. The system is designed for a 6 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (>12 m/s) salt streams and also recover up to half of the dynamic head. Cost estimates for a 1 GWe and 2 GWe reactor chamber are presented

Large packages for reactor decommissioning waste

International Nuclear Information System (INIS)

Price, M.S.T.

1991-01-01

This study was carried out jointly by the Atomic Energy Establishment at Winfrith (now called the Winfrith Technology Centre), Windscale Laboratory and Ove Arup and Partners. The work involved the investigation of the design of large transport containers for intermediate level reactor decommissioning waste, ie waste which requires shielding, and is aimed at European requirements (ie for both LWR and gas cooled reactors). It proposes a design methodology for such containers covering the whole lifetime of a waste disposal package. The design methodology presented takes account of various relevant constraints. Both large self shielded and returnable shielded concepts were developed. The work was generic, rather than specific; the results obtained, and the lessons learned, remain to be applied in practice

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1981-09-01

Research activities in the Division of Reactor Engineering in fiscal 1980 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 2. Attenuation of gamma rays. An example of shield's thickness calculation

International Nuclear Information System (INIS)

Lukaszek, W.; Kucypera, S.

1982-01-01

The semianalytic method was used for calculating the attenuation of gamma rays and the thickness of biological shield of graphite moderated reactor. A short description of computer code as well as the exemplary results of calculations are given. (A.S.)

Safety studies concerning nuclear power reactors

International Nuclear Information System (INIS)

Bailly, Jean; Pelce, Jacques

1980-01-01

The safety of nuclear installations poses different technical problems, whether concerning pressurized water reactors or fast reactors. But investigating methods are closely related and concern, on the one hand, the behavior of shields placed between fuel and outside and, on the other, analysis of accidents. The article is therefore in two parts based on the same plan. Concerning light water reactors, the programme of studies undertaken in France accounts for the research carried out in countries where collaboration agreements exist. Concerning fast reactors, France has the initiative of their studies owing to her technical advance, which explains the great importance of the programmes under way [fr

Neutron multiplication and shielding problems in pressurized water reactor spent fuel shipping casks

International Nuclear Information System (INIS)

Devillers, C.; Blum, P.

1977-01-01

To evaluate the degree of accuracy of computational methods used in the shield design of spent fuel shipping casks, comparisons have been made between biological dose-rate calculations and measurements at the surface of a cask carrying three pressurized water reactor fuel assemblies. Neutron dose-rate measurements made with the fuel-carrying region successively wet and dry are also used to derive an experimental value of the k/sub eff/ of the wet fuel assemblies. Results obtained by this method are shown to be consistent with criticality calculations, taking into account fuel depletion

Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

International Nuclear Information System (INIS)

Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

1978-01-01

An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

Reactor cavity streaming: the problem and engineered solutions

International Nuclear Information System (INIS)

Iotti, R.C.; Yang, T.L.; Rogers, W.H.

1979-01-01

Experience at operating pressurized water reactors has revealed that air gaps between the reactor vessel and the biological shield wall can provide paths for radiation streaming, which may prohibitively limit the accessibility required to areas in the containment during power operation, increase personnel exposure during shutdown, and cause radiation damage to equipment and cables located above the vessel. Several concepts of shield are discussed together with their predicted effectiveness. The analytical methods employed to determine the streaming magnitude and the shield effectiveness are also discussed and their accuracy is measured by comparison with actual measurement at an operating plant

Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

International Nuclear Information System (INIS)

Sato, Satoshi; Maegawa, Toshio; Yoshimatsu, Kenji; Sato, Koichi; Nonaka, Akira; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

2011-01-01

A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

Determination of boron in Jabroc wood used as a shielding material in nuclear reactors

International Nuclear Information System (INIS)

Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

2015-01-01

Jabroc are non-impregnated, densified wood laminates developed commercially for a wide range of industrial applications. Jabroc can be used with other neutron shielding materials such as Lead to form complex shielding structures. Its relative light weight and cleanliness in handling are additional features that make it a suitable candidate for the standard design of neutron shielding equipment. Jabroc can also be impregnated with Boron up to a maximum of 4% to be used in areas where Gamma radiation produced on Neutron capture reaches unacceptable dose rates. Boron impregnated Jabroc wood finds application in TAPS 3 and 4 as a shielding material for the Ion Chambers and the Horizontal Flux Units (HFU). The shielding property of this material is optimized by incorporating requisite amount of boron in wood. Boron content in this material has to be determined accurately prior to its use in the nuclear reactors. In this work a method was standardized to determine boron in Jabroc wood samples to check for conformance to specifications. The wood sample flakes were wetted with saturated barium hydroxide solution and dries under IR. The sample was ashed in a muffle furnace at 600℃ for 2 h

Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

International Nuclear Information System (INIS)

Kawahara, Toshio; Matsushita, Tadashi

1977-01-01

The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

Activity report of Reactor Physics Section - 1985

International Nuclear Information System (INIS)

John, T.M.

1986-01-01

This Activity Report contains brief summaries of different studies made in Reactor Physics Section during the year 1985. These are presented under the headings Nuclear Data Processing and Validation, Reactor Design and Analysis, Safety and Noise Analysis, Radiation Transport and Shielding, Reactor Physics Experiments and Statistical Physics. The work on nuclear data during this period comprises primarily of validation of data of 232 Th and 233 U as a part of participation in the Co-ordinated Research Programme (CRP) under IAEA research contract. The most significant event during 1985 at this centre has been the first criticality of FBTR (Fast Breeder Test Reactor), which was achieved on the 18th of October. Reactor Physics Section has played a key role in this event by carrying out the first approach to criticality with fuel loading in a safe manner and conducting some low power reactor physics experiments which are discussed. The studies made in the field reactor safety and shielding are also connected mainly with the FBTR problems in addition to some work on the PFBR (Prototype Fast Breeder Reactor) detailed design of which has been just started. Studies pertaining to the other two Co-ordinated Research Programmes (CRP) under IAEA contract, namely (1) on the comparative assessment of processing techniques for the analysis of sodium boiling noise detection and, (2) on the contribution of advanced reactors to energy supply have been continued during this year. At the end of this report, a list of publications made by the members of the section and also the sectional seminars held during this period is included. (author)

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

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

Commercial tokamak reactors with resistive toroidal field magnets

International Nuclear Information System (INIS)

Bombery, L.; Cohn, D.R.; Jassby, D.L.

1984-01-01

Scaling relations and design concepts are developed for commercial tokamak reactors that use watercooled copper toroidal field (TF) magnets. Illustrative parameters are developed for reactors that are scaled up in size from LITE test reactor designs, which use quasi-continuous copper plate magnets. Acceptably low magnet power requirements may be attainable in a moderate beta (β = 0.065) commercial reactor with a major radius of 6.2 m. The shielding thickness and magnet size are substantially reduced relative to values in commercial reactors with superconducting magnets. Operation at high beta (β = 0.14) leads to a reduction in reactor size, magnet-stored energy, and recirculating power. Reactors using resistive TF magnets could provide advantages of physically smaller devices, improved maintenance features, and increased ruggedness and reliability

Rectification of leak from upper aluminium thermal shield cooling water inlet line of Cirus reactor

International Nuclear Information System (INIS)

Bhatnagar, Anil; Joshi, N.S.; Kharpate, A.V.; Marik, S.K.

2006-01-01

During 1994, a small water leak was observed from the upper aluminium thermal shield of Cirus reactor. Detailed investigations revealed that the leakage was from the weld joint of one of the 1 1/4 inch NB Sch. 80 coolant inlet pipes connected to the upper aluminium thermal shield. The location of the leak was identified by monitoring the stabilised water level in the vertical inlet pipe under stagnant condition. The exact location was identified by installing an inflatable seal arrangement inside the leaky pipe and inflating the seal at different elevations to isolate the leaky location and ensuring that the leak was completely stopped. This location was about 15 feet below the operating floor of the reactor. The pipe was visually inspected with the help of a fibre-scope to assess the condition of the inner surface. Eddy current testing was also carried out for volumetric examination. This revealed one more localised flaw on the outer surface little above the leaky joint. A hollow plug, with expandable rings, having C-shaped cross section at both the ends and a straight portion in the middle to cover the defective region, was developed and qualified in a mock-up station after extensive trials. In view of the site constraints, a flexible hollow link assembly was engineered, for installing the plug remotely. The inner surface of the pipe was cleaned using an emery brush and a deburring tool. The plug was then installed covering the leak area and the rings were expanded by remote tightening. The shield was hydro-tested satisfactorily. (author)

SOLASE: a conceptual laser fusion reactor design

International Nuclear Information System (INIS)

Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

1977-12-01

The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute

Energy Technology Data Exchange (ETDEWEB)

Xavier, Larissa R.P.; Cardoso, Domingos D’Oliveira, E-mail: larissa.xavier@cnen.gov.br, E-mail: domingosoliveiralvr71@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Ferreira, Francisco José de Oliveira; Voi, Dante Luiz, E-mail: fferreira@ien.gov.br, E-mail: dante@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN's Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). (author)

Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute

International Nuclear Information System (INIS)

Xavier, Larissa R.P.; Cardoso, Domingos D’Oliveira; Ferreira, Francisco José de Oliveira; Voi, Dante Luiz

2017-01-01

Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN's Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). (author)

Operating manual for the Bulk Shielding Reactor

International Nuclear Information System (INIS)

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

Operating manual for the Bulk Shielding Reactor

International Nuclear Information System (INIS)

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

Operating manual for the Bulk Shielding Reactor

Energy Technology Data Exchange (ETDEWEB)

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

Operating manual for the Bulk Shielding Reactor

Energy Technology Data Exchange (ETDEWEB)

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

International Nuclear Information System (INIS)

Pattinson, A.

2003-01-01

The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

International Nuclear Information System (INIS)

1986-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

Repairing liner of the reactor

International Nuclear Information System (INIS)

Aguilar H, F.

2001-07-01

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Acoustic emission technique for leak detection in an end shield of a pressurized heavy water reactor

International Nuclear Information System (INIS)

Kalyanasundaram, P.; Jayakumar, T.; Raj, B.

1989-01-01

This paper discusses the successful application of the Acoustic Emission Technique (AET) for detection and location of leak paths present on the inaccessible side of an end shield of a Pressurized Heavy Water Reactor (PHWR). The methodology was based on the fact that air and water leak AE signals have different characteristic features. Baseline data was generated from a sound end-shield of a PHWR for characterizing the background noise. A mock up end-shield system with saw cut leak paths was used to verify the validity of the methodology. It was found that air leak signals under pressurisation (as low as 3 psi) could be detected by frequency domain analysis. Signals due to air leaks from various locations of a defective end-shield were acquired and analysed. It was possible to detect and locate leak paths. Presence of detected leak paths were further confirmed by alternate test. (orig.)

ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

International Nuclear Information System (INIS)

1988-06-01

This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

Air conditioning device for reactor buildings

International Nuclear Information System (INIS)

Kikuchi, Shiro.

1982-01-01

Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1975-01-01

Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

International Nuclear Information System (INIS)

El-Guebaly, L.A.; Sviatoslavsky, I.N.

1987-01-01

The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m 2 neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10 19 n/cm 2 . In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H 2 O/LiNO 3 layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is 0 C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of ∼2 were found at the hot spots. 2 refs., 6 figs., 2 tabs

Formulary for neutron propagation in sodium-steel media for the fast reactor shields

International Nuclear Information System (INIS)

Bouteau, F.; Caumette, P.; Khairallah, A.; Oceraies, Y.; Devillers, C.

1975-01-01

The simplified calculational tool (''formulary'') for neutron propagation in the shields of fast reactors, being developed at CEA, has two objectives: to reduce the cost of the major part of design calculations, without a significant loss of accuracy; to facilitate the adjustment of the calculational tool with the results of the program of integral propagation experiments, which is conducted in parallel with the development of the calculational method. The version 0 (i.e. before any adjustment) of the formulary and a first test of its validity as compared to the results of integral measurements are presented [fr

Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

International Nuclear Information System (INIS)

1987-08-01

Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

Bench-mark experiments to study the neutron distribution in a heterogeneous reactor shielding

International Nuclear Information System (INIS)

Bolyatko, V.V.; Vyrskij, M.Yu.; Mashkovich, V.P.; Nagaev, R.Kh.; Prit'mov, A.P.; Sakharov, V.K.; Troshin, V.S.; Tikhonov, E.G.

1981-01-01

The bench-mark experiments performed at the B-2 facility of the BR-10 reactor to investigate the spatial and energy neutron distributions are described. The experimental facility includes the neutron beam channel with a slide, a mo shielding composition investigated consisted of sequential layers of steel (1KH18N9T) and graphite slabs. The neutron spectra were measured by activation method, a set of treshold and resonance detectors having been used. The detectors made it possible to obtain the absolute neutron spectra in the 1.4 eV-10 MeV range. The comparison of calculations with the results of the bench-mark experiments made it possible to prove the neutron transport calculational model realized in the ROZ-9 and ARAMAKO-2F computer codes and evaluate the validity of the ARAMAKO constants for the class of shielding compositions in question [ru

HYLIFE-II reactor chamber mechanical design: Update

International Nuclear Information System (INIS)

House, P.A.

1992-01-01

Mechanical design features of the reactor chamber for the HYLIFE-II inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams (Li 2 BeF 4 ) are used for shielding and blast protection of the chamber walls. The system is designed for a 6 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (17 m/s) salt streams and also recover up to half of the dynamic head. Cost estimates for a 1 GW e and 2 GW e reactor chamber are presented

Need of Reactor Dosimetry Preservation

International Nuclear Information System (INIS)

Ilieva, Krassimira

2011-01-01

The nuclear safety requirements and philosophy have changed by the development of new nuclear systems and this imposes special research and development activity. Reactor dosimetry which is applied for determination of neutron field parameters and neutron flux responses in different regions of the reactor system plays an important role in determining of radiation exposure on reactor system elements as reactor vessel, internals, shielding; dose determination for material damage study; for conditioning of irradiation; dose determination for medicine and industry application; induced activity determination for decommissioning purposes. The management of nuclear knowledge has emerged as a growing challenge in recent years. The need to preserve and transfer nuclear knowledge is compounded by recent trends such as ageing of the nuclear workforce, declining student numbers in nuclear related fields, and the threat of losing accumulated nuclear knowledge. (author)

Scyllac fusion test reactor design

International Nuclear Information System (INIS)

Dudziak, D.J.; Gerstl, S.A.; Houck, D.L.; Jalbert, R.A.; Krakowski, R.A.; Linford, R.K.; McDonald, T.E.; Rogers, J.D.; Thomassen, K.I.

1975-01-01

A general design of the system is given. The implosion heating and compression systems (METS) are described. Tritium handling, shielding and activation of the reactor, and safety and environmental aspects are discussed

Modeling the PUSPATI TRIGA Reactor using MCNP code

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Mark Dennis Usang; Naim Syauqi Hamzah; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

2012-01-01

The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)

Shielding computations for solution transfer lines from Analytical Lab to process cells of Demonstration Fast Reactor Plant (DFRP)

International Nuclear Information System (INIS)

Baskar, S.; Jose, M.T.; Baskaran, R.; Venkatraman, B.

2018-01-01

The diluted virgin solutions (both aqueous and organic) and aqueous analytical waste generated from experimental analysis of process solutions, pertaining to Fast Breeder Test Reactor (FBTR) and Prototype Fast Breeder Reactor (PFBR), in glove boxes of active analytical Laboratory (AAL) are pumped back to the process cells through a pipe in pipe arrangement. There are 6 transfer lines (Length 15-32 m), 2 for each type of transfer. The transfer lines passes through the area inside the AAL and also the operating area. Hence it is required to compute the necessary radial shielding requirement around the lines to limit the dose rates in both the areas to the permissible values as per the regulatory requirement

First preliminary design of an experimental fusion reactor

International Nuclear Information System (INIS)

1977-09-01

A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

International Nuclear Information System (INIS)

Salahuddin, A.; Israr, M.; Hussain, M.

2002-01-01

Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

Measurement of the thermal neutron self shielding coefficient in the Syrian miniature neutron source reactor inner irradiation site using the dy soils

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2007-01-01

Measurement of the thermal self shielding coefficient ( Gth ) in the Syrian Miniature Neutron Source Reactor (MNSR) inner irradiation site using Dy foils is presented in this paper. The thermal self shielding coefficient is measured as a function of the foil thickness or numbers. The mathematical equation which calculates the average relative radioactivity (Bq/g) versus the foil number is found as well.

Remote maintenance design for Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Tachikawa, K.; Iida, H.; Nishio, S.; Tone, T.; Aota, T.; Iwamoto, T.; Niikura, S.; Nishizawa, H.

1984-01-01

Design of Fusion Experimental Reactor, FER, has been conducted by Japan Atomic Energy Research Institute (JAERI) since 1981. Two typical reactors can be classified in general from the viewpoints of remote maintenance among four design concepts of FER. In the case of the type 1 FER, the torus module consists of shield structure and blanket, and the connective joints between toruses provided at the outer region of the reactor. As for the type 2 FER, the shield structure is joined with the vacuum cryostat, and only the blanket module is allowed to move, but connection between toruses are located in the inner region of the reactor. Comparing type 1 with type 2 FER, this paper describes on the remote maintenance of FER including reactor configurations, work procedures, remote systems/equipments, repairing facility and future R and D problems. Reviewing design studies and investigation for the existing robotics technologies, R and D for FER remote maintenance technology should be performed under the reasonable long-term program. The main items of remote technology required to start urgently are multi-purpose manipulator system with performance of dextrousity, tele-viewing system which reduces operator fatigue and remote tests for commercially available components

Second preliminary design of JAERI experimental fusion reactor (JXFR)

International Nuclear Information System (INIS)

Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

1979-06-01

Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

HYLIFE-II reactor chamber mechanical design

International Nuclear Information System (INIS)

House, P.A.

1992-01-01

Mechanical design features of the reactor chamber for the HYLIFE-11 inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams are used for shielding and blast protection. The system is designed for an 8 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (20 m/s) salt streams and also recover up to half of the dynamic head

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

Nuclear reactor can detonate like an atomic bomb: yes or no

International Nuclear Information System (INIS)

Martin, D.

1974-01-01

The fission process in a nuclear weapon and in power reactor are compared. The operation of a power reactor is described and the various protective systems and shielding devices are mentioned. It is shown that it is not possible for a nuclear explosion to occur in a power reactor

Plasma engineering analysis of a small torsatron reactor

International Nuclear Information System (INIS)

Lacatski, J.T.; Houlberg, W.A.; Uckan, N.A.

1985-10-01

This study examines the plasma physics and reactor engineering feasibility of a small, medium aspect ratio, high-beta, l = 2, D-T torsatron power reactor, based on the magnetic configuration of the Advanced Toroidal Facility, Oak Ridge National Laboratory. Plasma analyses are performed to assess whether confinement in a small, average radius plasma is sufficient to yield an ignited or high-Q driven device. Much of the physics assessment focuses on an evaluation of the radial electric field created by the nonambipolar particle flux. Detailed transport simulations are done with both fixed and self-consistent evolution of the radial electric field. Basic reactor engineering considerations taken into account are neutron wall loading, maximum magnetic field at the helical coils, coil shield thickness, and tritium breeding blanket-shield thickness

US DOE Idaho national laboratory reactor decommissioning

International Nuclear Information System (INIS)

Szilagyi, Andrew

2012-01-01

, asbestos and mercury among others. Each reactor required isolation in order to be removed. Due to activated metal within the reactor vessels, dose rates above the cores ranged from 50 R/hr to 1200 R/hr. Subsequent dose rates outside the vessels varied from 60 mR/hr to greater than 50 R/hr. Due to the elevated dose rates, the project team decided to fill the ETR and MTR reactor vessels with grout to a level above the core region to reduce dose. To remove the ETR reactor, access to the support shoes was required. These shoes were encased in the high density concrete biological shield approximately 8' below grade. The project team used explosives to remove the biological shield. The demolition had to be controlled to prevent damaging the reactor vessel and to limit the seismic impact on a nearby operating reactor. Upon completion of the blast, the concrete was removed exposing the support shoes for the vessel. Two reactor buildings (ETR and PBF) had to be removed to accommodate lifting systems for the reactor vessels. Two reactors (PBF and MTR) were removed via mobile cranes, two reactors were sized and removed in pieces (ZPPR and MTR), and ETR reactor, due to its weight, was removed via a twin gantry lifting system

Spatial distribution of reactor radiation around the horizontal experimental hole of the RA Reactor at Vinca; Prostorna raspodela reaktorskog zracenja oko horizontalnog eksperimentalnog kanala reaktora RA u Vinci

Energy Technology Data Exchange (ETDEWEB)

Ninkovic, M; Paligoric, D; Vujisic, B [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1964-10-15

A survey is given of the measurement of the spatial distribution of thermal and fast neutrons and gamma radiations inside and around the reactor radiation beam which is carried through the experimental channel outside the reactor biological shield thus being a potential danger for the personnel in the reactor hall. The results obtained are used to determine the optimal distribution of the shielding elements around the free radiation beam. The activation method was used to determine the distribution of slow and fast neutrons. The detectors were Au{sup 197} and In {sup 115} for slow, and the S{sup 32} (n,p) P{sup 32} reaction for fast neutrons. The relative distribution of slow and fast neutrons along the beam axis in the space outside the reactor, and evaluation of the absolute value of the flux at the place where the beam comes out from the biological shield are given (author)

Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

International Nuclear Information System (INIS)

Parkes, Alison; Saratsopoulos, John; Deura, Michael; Kenny, Pat

2008-01-01

The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-01-01

Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

Activity report of Reactor Physics Division - 1988

International Nuclear Information System (INIS)

Keshavamurthy, R.S.

1989-01-01

This report highlights the progress of activities carried out during the year 1988 in Reactor Physics Division in the form of brief summaries. The topics are organised under the following subject categories:(1) nuclear data evaluation , processing and validation, (2) core physics and analysis, (3) reactor kinetics and safety analysis, (4) noise analysis and (5) radiation transport and shielding. List of publications by the members of the Division and the Reactor Physics Seminars held during the year 1988, is included at the end of report. (author). refs., figs., tabs

Benchmark shielding calculations for the NEACRP [Nuclear Energy Agency-Committee on Reactor Physics] Working Group on shielding assessment of transportation packages

International Nuclear Information System (INIS)

Broadhead, B.L.; Brady, M.C.; Parks, C.V.

1990-11-01

In 1985, the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP) established a working group on shielding assessment of transportation packages. Following the initial distribution of a set of six problems, discussions were held at the Organization for Economic Cooperation and Development (OECD) Headquarters in Paris, France, in June/July 1986, May 1988, and February/March 1990. The US contribution to the working group is documented in this report. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask. Among the effects reported here are the performance of multiple cross-section sets, the comparison of several source generation codes, and multidimensional versus one-dimensional (1-D) analyses. 18 refs., 16 figs., 33 tabs

Reactor container

International Nuclear Information System (INIS)

Ichiki, Tadaharu; Saba, Kazuhisa.

1979-01-01

Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

International Nuclear Information System (INIS)

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

2001-01-01

The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

Maintenance features of the Compact Ignition Tokamak fusion reactor

International Nuclear Information System (INIS)

Spampinato, P.T.; Hager, E.R.

1987-01-01

The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs

Tokamak engineering test reactor

International Nuclear Information System (INIS)

Conn, R.W.; Jassby, D.L.

1975-07-01

The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

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)

Leak detection device for reactor coolant

International Nuclear Information System (INIS)

Oshima, Koichiro.

1990-01-01

In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

New developments in resonant mixture self-shielding treatment with Apollo code and application to Jules Horowitz reactor core calculation

International Nuclear Information System (INIS)

Coste-Delclaux, M.; Aggery, A.; Huot, N.

2005-01-01

APOLLO2 is a modular multigroup transport code developed by Cea in Saclay. Until last year, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Last year, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The calculations of a simplified Jules Horowitz reactor using a Monte-Carlo code (TRIPOLI4) as a reference and APOLLO2 in its standard and improved versions, show that, as far as the effective multiplication factor is concerned, the mixture treatment does not bring an improvement, because the new treatment suppresses compensation between the reaction rate discrepancies. The discrepancy of 300 pcm that appears with the reference calculation is in accordance with the technical specifications of the Jules Horowitz reactor

Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics

International Nuclear Information System (INIS)

Santos Bastos, W. dos

1995-01-01

These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods

Water chemistry management of research reactor in JAERI

Energy Technology Data Exchange (ETDEWEB)

Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

Tokamak fusion reactors with less than full tritium breeding

International Nuclear Information System (INIS)

Evans, K. Jr.; Gilligan, J.G.; Jung, J.

1983-05-01

A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

Nuclear reactor constructions

International Nuclear Information System (INIS)

Baddley, A.H.

1981-01-01

A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

Decommissioning and re-utilization of the Musashi Reactor

International Nuclear Information System (INIS)

Tomio Tanzawa; Nobukazu Iijima; Norikazu Horiuchi; Tadashi Yoshida; Tetsuo Matsumoto; Naoto Hagura; Ryouhei Kamiya

2008-01-01

The Musashi Institute of Technology Research Reactor (the Musashi Reactor) is a TRIGA-? with maximum thermal power of 100 kW. The decommissioning was decided in May, 2003. The reactor facility is now under decommissioning. The phased decommissioning was selected. Phase 1 consists of permanent shutdown of the reactor and stopping the operational functions, and transportation of the spent nuclear fuels. After completion of the transportation, the reactor facility is characterized as the storage of low level radioactive materials. This is phase 2. Activities of phase 1 were completed and the facility is now under phase 2. Activities of phase 3 consist of dismantling the reactor tank and the shielding, and delivering the radioactive waste to a waste disposal facility. The phase 3 will be started on condition that the undertaking of the waste disposal for research reactors will be established. On the other hand, re-utilization of the facility has being studied, and 'realistic' reactor simulator was turned out by utilizing the reactor installations such as control rod drive and operation console. (authors)

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Painter, S.L.

1992-12-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

Temperature distribution due to the heat generation in nuclear reactor shielding

International Nuclear Information System (INIS)

Torres, L.M.R.

1985-01-01

A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author) [pt

ATFSR: a small torsatron reactor

International Nuclear Information System (INIS)

Houlberg, W.A.; Lacatski, J.T.; Uckan, N.A.

1985-01-01

A small (average minor radius anti a approx. = 1 m), moderate-aspect-ratio torsatron reactor based on the Advanced Toroidal Facility (ATF) is proposed as a starting point for improved stellarator reactor designs. The major limitation of the compact size is the lack of space under the helical coils for the blanket and shield. Neoclassical confinement models for helically trapped particles show that a large electric potential (radial electric field) is necessary to achieve ignition in a device of this size, although high-Q operation is still attainable with more modest potentials

SOLASE: a conceptual laser fusion reactor design

International Nuclear Information System (INIS)

Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

1977-12-01

The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion ractors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

Reactor dosimetry knowledge preservation and development

International Nuclear Information System (INIS)

Ilieva, K.D.; Belousov, S.I.; Mitev, M.R.

2010-01-01

The nuclear safety requirements and philosophy have changed by the development of new nuclear systems and this imposes special research and development activity. Reactor dosimetry which is applied for determination of neutron field parameters and neutron flux responses in different regions of the reactor system plays an important role in determining of radiation exposure on reactor system elements as reactor vessel, internals, shielding; dose determination for material damage study; in dose determination and conditioning of irradiation for medicine and industry application as well as in induced activity determination for decommissioning purposes. The management of nuclear knowledge has emerged as a growing challenge in recent years. The need to preserve and transfer nuclear knowledge is compounded by recent trends such as ageing of the nuclear workforce, declining student numbers in nuclear related fields, and the threat of losing accumulated nuclear knowledge. (authors)

Reactor refurbishment options for a changing climate

Energy Technology Data Exchange (ETDEWEB)

McNeish, D. [Bruce Power, Tiverton, Ontario (Canada)

2012-07-01

As the industry looks ahead to another generation of reactor refurbishment, it is acknowledged that the traditional way of Retubing a reactor is a daunting prospect for our investors and stakeholders. Innovations are required to mitigate the long downtime and large one-time investment associated with previous reactor refurbishments. These can take the shape of improvements to the Retube processes or by fundamentally changing the approach, e.g., calandria/shield tank replacement or partial Retubes. This session presents technical challenges that utilities need help resolving to arrive at a more attractive reactor refurbishment model. This includes issues related to calandria vessel fitness-for-service, the fuel channel replacement process, the feeder replacement process, life extension of fuel channels and feeders and complexities involving interfacing systems. (author)

Reactor refurbishment options for a changing climate

International Nuclear Information System (INIS)

McNeish, D.

2012-01-01

As the industry looks ahead to another generation of reactor refurbishment, it is acknowledged that the traditional way of Retubing a reactor is a daunting prospect for our investors and stakeholders. Innovations are required to mitigate the long downtime and large one-time investment associated with previous reactor refurbishments. These can take the shape of improvements to the Retube processes or by fundamentally changing the approach, e.g., calandria/shield tank replacement or partial Retubes. This session presents technical challenges that utilities need help resolving to arrive at a more attractive reactor refurbishment model. This includes issues related to calandria vessel fitness-for-service, the fuel channel replacement process, the feeder replacement process, life extension of fuel channels and feeders and complexities involving interfacing systems. (author)

Small reactor power systems for manned planetary surface bases

Energy Technology Data Exchange (ETDEWEB)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

Small reactor power systems for manned planetary surface bases

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

International Nuclear Information System (INIS)

1988-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

Assessment of beam tube performance for the maple research reactor

International Nuclear Information System (INIS)

Lee, A.G.

1986-06-01

The MAPLE research reactor is a versatile new research facility that can be adapted to meet the requirements of a variety of reactor applications. A particular group of reactor applications involves the use of beams of radiation extracted from the reactor core via tubes that penetrate through the biological shield and terminate in the reflector surrounding the fuelled core. An assessment is given of the neutron and gamma radiation fields entering beam tubes that are located radially or tangentially with respect to the core

Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

International Nuclear Information System (INIS)

Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

1985-01-01

A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

International Nuclear Information System (INIS)

Lawrence, R.D.; Dorning, J.J.

1980-01-01

A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method

Conceptual design study of fusion experimental reactor (FY 86 FER)

International Nuclear Information System (INIS)

Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

1987-09-01

This report describes the results of the investigation on critical issues of FY 86 FER reactor configuration/structure design. Accuracy evaluation of shielding calculation and crack growth prediction of first wall and divertor based on the elastic-plastic fracture mechanics were performed. Further, optimization of shield configuration, graphite first wall armor and flexifility of reactor were investigated to support future design work. Feasibilities of innovative ideas were also examined, such as the ripple insert effect and the application of shape memory alloys. (author)

Ground testing of an SP-100 prototypic reactor

International Nuclear Information System (INIS)

Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

1988-01-01

SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

Calculation of releases of radioactive materials in gaseous effluents from nuclear-powered merchant ships (NMS-GEFF code)

International Nuclear Information System (INIS)

Cardile, F.P.; Bangart, R.L.; Collins, J.T.

1978-06-01

The Intergovernmental Maritime Consultative Organization IMCO) is currently preparing guidelines concerning the safety of nuclear-powered merchant ships. An important aspect of these guidelines is the determination of the releases of radioactive material in effluents from these ships and the control exercised by the ships over these releases. To provide a method for the determination of these releases, the NRC staff has developed a computerized model, the NMS-GEFF Code, which is described in the following chapters. The NMS-GEFF Code calculates releases of radioactive material in gaseous effluents for nuclear-powered merchant ships using pressurized water reactors

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Possibilities of miniaturizing the TRIGA-reactor

International Nuclear Information System (INIS)

Bobleter, O.; Brunner, P.; Schachner, H.

1976-01-01

It is proposed to decrease the depth of the TRIGA pool in cases where the construction of the normal-sized pool causes difficulty. The loss of shielding in the vertical direction will be compensated by lead and lead glass. The influence of these changes in design on the reactor components (control rods, instrumentation, neutron beam tubes, pneumatic system, etc.) is discussed. The experimental part of the work concerns the irradiation of lead glasses with varying contents of lead and cerium, which was carried out in the pool at different distances from the TRIGA core. The advantages of a possible reduction in size of the TRIGA reactor by using lead and lead glass as shielding are compared with the main disadvantages of these materials (darkening of the glass and high prices). (author)

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)

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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)

Nuclear Reactor RA Safety Report, Vol. 13, Causes of possible accidents

International Nuclear Information System (INIS)

1986-11-01

This volume includes the analysis of possible accidents on the RA research reaktor. Any unwanted action causing decrease of integrity of any of the reactor safety barriers is considered to be a reactor accident. Safety barriers are: fuel element cladding, reactor vessel, biogical shield, and reactor building. Reactor accidents can be classified in four categories: (1) accidents caused by reactivity changes; (2) accidents caused by mis function of the cooling system; (3) accidents caused by errors in fuel management and auxiliary systems; (4) accidents caused by natural or other external disasters. The analysis of possible causes of reactor accidents includes the analysis of possible impacts on the reactor itself and the environment [sr

Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson

1996-01-01

A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

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

Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

International Nuclear Information System (INIS)

1989-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Structure of pool in reactor building

International Nuclear Information System (INIS)

Yokoyama, Shigeki.

1997-01-01

Shielding walls made of iron-reinforced concrete having a metal liner including two body walls rigidly combined to the upper surface of a reactor container are disposed at least to one of an equipment pool or spent fuel storage pool in a reactor building. A rack for temporarily placing an upper lattice plate is detachably attached at least above one of a steam dryer or a gas/liquid separator temporarily placed in the temporary pool, and the height from the bottom portion to the upper end of the shielding wall is determined based on the height of an upper lattice plate temporary placed on the rack and the water depth required for shielding radiation from the upper lattice plate. An operator's exposure on the operation floor can be reduced by the shielding wall, and radiation dose from the spent fuels is reduced. The increase of the height of a pool guarder enhances bending resistance as a ceiling. In addition, the total height of them is made identical with the depth of the spent fuel storage pool thereby enabling to increase storage area for spent fuels. (N.H.)

Lower activation materials and magnetic fusion reactors

International Nuclear Information System (INIS)

Conn, R.W.; Bloom, E.E.; Davis, J.W.; Gold, R.E.; Little, R.; Schultz, K.R.; Smith, D.L.; Wiffen, F.W.

1984-01-01

Radioactivity in fusion reactors can be effectively controlled by materials selection. The detailed relationship between the use of a material for construction of a magnetic fusion reactor and the material's characteristics important to waste disposal, safety, and system maintainability has been studied. The quantitative levels of radioactivation are presented for many materials and alloys, including the role of impurities, and for various design alternatives. A major outcome has been the development of quantitative definitions to characterize materials based on their radioactivation properties. Another key result is a four-level classification scheme to categorize fusion reactors based on quantitative criteria for waste management, system maintenance, and safety. A recommended minimum goal for fusion reactor development is a reference reactor that (a) meets the requirements for Class C shallow land burial of waste materials, (b) permits limited hands-on maintenance outside the magnet's shield within 2 days of a shutdown, and (c) meets all requirements for engineered safety. The achievement of a fusion reactor with at least the characteristics of the reference reactor is a realistic goal. Therefore, in making design choices or in developing particular materials or alloys for fusion reactor applications, consideration must be given to both the activation characteristics of a material and its engineering practicality for a given application

International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Blevins, J.D.; Stasko, R.R.

1989-09-01

An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

Development of a shielded ion microprobe analyzer (SIMA) and its application to fast reactor fuel elements

International Nuclear Information System (INIS)

Yuji, E.; Junji, K.; Sadamu, Y.; Toshiyuki, I.

1983-01-01

A shielded ion microprobe analyzer for elemental and isotopic analyses of irradiated fast reactor fuel and fuel component has been developed and installed in an alpha-gamma hot cell. Radiation shielding of the equipment ensures the radiation dose of -7 C/kg) for 5 Ci (1.85 x 10 11 Bq) of a 60 Co source. Hot samples can be automatically transferred from the cell to the sample chamber of the analyzer. Contamination inside the equipment through sputtering of the radioactive materials can be reduced with a special device. Distribution and migration of fission products, such as 137 Cs, 138 Ba, and 90 Sr, and of fissile materials, such as 235 U and 239 Pu in irradiated mixed-oxide fuel, and isotopic ratios of the elements can be obtained very precisely and quickly

Development of a shielded ion microprobe analyzer (SIMA) and its application to fast reactor fuel elements

International Nuclear Information System (INIS)

Enokido, Y.; Itaki, T.; Komatsu, J.; Yamanouchi, S.

1983-01-01

A shielded ion microprobe analyzer for elemental and isotopic analyses of irradiated fast reactor fuel and fuel component has been developed and installed in an alpha-gamma hot cell. Radiation shielding of the equipment ensures the radiation dose of -7 C/kg) for 5 Ci (1.85 X 10 11 Bq) of a 60 Co source. Hot samples can be automatically transferred from the cell to the sample chamber of the analyzer. Contamination inside the equipment through sputtering of the radioactive materials can be reduced with a special device. Distribution and migration of fission products, such as 137 Cs, 138 Ba, and 90 Sr, and of fissile materials, such as 235 U and 239 Pu in irradiated mixed-oxide fuel, and isotopic ratios of the elements can be obtained very precisely and quickly

Small intrinsically safe reactor implications

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1985-01-01

Reviewing the history of nuclear power, it is found that peaceful uses of nuclear power are children of the war-like atom. Importance of special growth in a shielded environment is emphasized to exploit fully the advantages of nuclear power. Nuclear power reactors must be safe for their assimilation into society from the points of view of both technology and social psychology. ISR/ISER is identified as a missing link in the development of nuclear power reactors from this perspective and advocated for international development and utilization, being unleashed from the concerns of politicization, safety, and proliferation

Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

International Nuclear Information System (INIS)

Schweizer, Fernando Lage Araújo

2014-01-01

The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

Procedure for operating a heavy water cooled power reactor

International Nuclear Information System (INIS)

Rau, P.; Kumpf, H.

1981-01-01

Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

A Compact Quasi-axisymmetric Stellarator Reactor

International Nuclear Information System (INIS)

Ku, L.P.

2003-01-01

We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1962-09-15

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined.

Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

International Nuclear Information System (INIS)

1991-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

International Nuclear Information System (INIS)

1992-08-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Space reactor electric systems: system integration studies, Phase 1 report

International Nuclear Information System (INIS)

Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

1983-01-01

This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

International Nuclear Information System (INIS)

Belousov, S.; Antonov, S.; Ilieva, K.

1997-01-01

The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

2008-01-01

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

2009-01-01

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

Research reactor RB, technical characteristics and experimental possibilities

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1978-01-01

Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) [sr

System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Lewis, B.R.; Klein, A.C.; Pawlowski, R.A.

1993-01-01

Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range

The advanced MAPLE reactor concept

International Nuclear Information System (INIS)

Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

1989-01-01

In Canada the need for advanced neutron sources has long been recognized. During the past several years Atomic Energy of Canada Limited (AECL) has been developing the new MAPLE multipurpose reactor concept. To date, the MAPLE program has focused on the development of a modest-cost multipurpose medium-flux neutron source to meet contemporary requirements for applied and basic research using neutron beams, for small-scale materials testing and analysis and for radioisotope production. The basic MAPLE concept incorporates a compact light-water cooled and moderated core within a heavy water primary reflector to generate strong neutron flux levels in a variety of irradiation facilities. In view of renewed Canadian interest in a high-flux neutron source, the MAPLE group has begun to explore advanced concepts based on AECL's experience with heavy water reactors. The overall objective is to define a high-flux facility that will support materials testing for advanced power reactors, new developments in extracted neutron-beam applications, and/or production of radioisotopes. The design target is to attain performance levels of HFR-Grenoble, HFBR, HFIR in a new heavy water-cooled, -moderated,-reflected reactor based on rodded LEU fuel. Physics, shielding, and thermohydraulic studies have been performed for the MAPLE heavy water reactor. 14 refs., 4 figs., 1 tab

Heat pipe nuclear reactor for space power

Science.gov (United States)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

The 33 years of research reactors in JAERI

International Nuclear Information System (INIS)

1990-11-01

The development and utilization of atomic energy in Japan began with the installation of JRR-1 reactor which attained the criticality in August, 1957, thus the third fire was lighted for the first time in Japan. JRR-2 was constructed as a full scale versatile research reactor, which attained the criticality in October, 1960, and since 1962, it has accomplished the role of the reactor for joint utilization. JRR-3 is the first reactor made in Japan by concentrating Japanese technologies in it, to develop and improve Japanese atomic energy technology. It attained the criticality in September, 1962, and has been used as a versatile research reactor. In 1960, Research Reactor Management Department was founded. JRR-4 was constructed as the research reactor for shielding for developing a nuclear-powered ship, which attained the criticality in January, 1965. The first hot laboratory in Japan for carrying out the post-irradiation test on the fuel and materials irradiated in these research reactors was installed in 1961. The JRR-1 was stopped in September, 1968, and is used as the commemorative exhibition hall. The JRR-3 was reconstructed, and attained the criticality in March, 1990, using 20 % enriched uranium fuel. The course of the research reactors for 33 years is reported. (K.I.)

The shielding calculation for the CN guide shielding assembly in HANARO

International Nuclear Information System (INIS)

Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

2006-01-01

The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

Status report on the Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D ampersand D) Project

International Nuclear Information System (INIS)

Sears, L.; Garlock, G.; Mencarelli, R.; Fellhauer, C.

1994-01-01

ALARON Corporation is under contract, to Argonne National Laboratory - East (ANL-E), to complete the decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR). The project, begun, in 1986 by ANL-E personnel, is projected to be completed by the end of 1994. The final phase of work was awarded to ALARON in December 1993 with the scope of work including the disassembly and removal of all remaining reactor internals, the reactor vessel, the lead bio-shield, the core liner, and the activated portion of the concrete bio-shield. This paper discusses the work undertaken beginning in January 1994 and continuing through July 1994. During this period the required pre-mobilization documentation was prepared and approved, mobilization was completed, and the reactor internals, reactor vessel, lead bio-shield and core liner were removed. The paper will compare the planned schedule to the actual schedule, discuss problems encountered, review volume reduction techniques and health and safety issues including radiological aspects of the project

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Certifying the decommissioned Shippingport reactor vessel for transport

International Nuclear Information System (INIS)

Towell, R.H.

1990-01-01

The decommissioned Shippingport reactor pressure vessel with its concentric neutron shield tank was shipped to Hanford, WA as part of the effort to restore the Shippingport Station to its original condition. The metal walls of the reactor vessel had become radioactive from neutron bombardment while the reactor was operating so it had to be shipped under the regulations for transporting radioactive material. Because of the large amount of radioactivity in the walls, 16,467 Curies, and because the potentially dispersible corrosion layer on the inner walls of both tanks was also radioactive, the Shippingport reactor vessel was transported under the most stringent of the regulations, those for a type B package. Compliance with the packaging regulations was confirmed via independent analysis by the staff of the Department of Energy certifying official and the Shippingport reactor vessel was shipped under DOE Certificate of Compliance USA/9515/B(U)

Proceedings of the symposium on the physics and technology of reactors

International Nuclear Information System (INIS)

1993-01-01

The symposium aimed at providing the opportunity for promoting the subject and for developing the human resources in this important field in the Arab States. The symposium included 32 lectures on the following topics related to research reactors: design and development, training and operation, calculations of reactor parameters, nuclear reactions dynamics and control, reactor physics, neutron pyhsics, neutron activation analysis, in-core reactor radiation protection and shielding calculations. The lectures of the symposium were distributed over 7 sessions. An additional session was held by all participants for open discussion and recommendations

Engineering design of advanced marine reactor MRX

International Nuclear Information System (INIS)

1997-10-01

JAERI has studied the design of an advanced marine reactor (named as MRX), which meets requirements of the enhancement of economy and reliability, by reflecting results and knowledge obtained from the development of N.S. Mutsu. The MRX with a power of 100 MWt is intended to be used for ship propulsion such as an ice-breaker, container cargo ship and so on. After completion of the conceptual design, the engineering design was performed in four year plan from FY 1993 to 1996. (1) Compactness, light-weightiness and simplicity of the reactor system are realized by adopting an integral-type PWR, i.e. by installing the steam generator, the pressurizer, and the control rod drive mechanism (CRDM) inside the pressure vessel. Because of elimination of the primary coolant circulation pipes in the MRX, possibility of large-scale pipe break accidents can be eliminated. This contributes to improve the safety of the reactor system and to simplify the engineered safety systems. (2) The in-vessel type CRDM contributes not only to eliminate possibilities of rod ejection accidents, but also to make the reactor system compact. (3) The concept of water-filled containment where the reactor pressure vessel is immersed in the water is adopted. It can be of use for emergency core cooling system which maintains core flooding passively in case of a loss-of-coolant accident. The water-filled containment system also contributes essentially light-weightness of the reactor system since the water inside containment acts as a radiation shield and in consequence the secondary radiation shield can be eliminated. (4) Adoption of passive decay heat removal systems has contributed in a greater deal to simplification of the engineered safety systems and to enhancement of reliability of the systems. (5) Operability has been improved by simplification of the whole reactor system, by adoption of the passive safety systems, advanced automatic operation systems, and so on. (J.P.N.)

FBR type reactors

International Nuclear Information System (INIS)

Otsuka, Masaya; Yamakawa, Masanori; Goto, Tadashi; Ikeuchi, Toshiaki; Yamaki, Hideo.

1986-01-01

Purpose: To prevent thermal deformation and making the container compact by improving the cooling performance of main container walls. Constitution: A pipeway is extended from a high pressure plenum below the reactor core and connected to the lower side of the flow channel at the inside of a thermal shielding layer disposed to the inside of the main container wall. Low pressure sodium sent from the low temperature plenum into the high pressure plenum is introduced to the pipeway, caused to uprise in the inside flow channel, then turned for the direction, caused to descend in the outer side flow channel between the main container and the inside flow channel and then returned to the low temperature plenum. A heat insulating layer disposed with argon gas is installed to the inside of the flow channel to reduce the temperature change applied upon reactor scram. An annular linear induction pump capable of changing the voltage polarity is disposed at the midway of the pipeway and the polarity is switched such that the direction of flow of the liquid sodium is exerted as a braking force upon rated operation, whereas exerted as a pumping force upon reactor scram. (Sekiya, K.)

Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

International Nuclear Information System (INIS)

Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

2010-01-01

Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Experience in utilizing research reactors in Yugoslavia

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

1972-07-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Experience in utilizing research reactors in Yugoslavia

International Nuclear Information System (INIS)

Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

1972-01-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Safety benefits from CANDU reactor replacement - a case study

International Nuclear Information System (INIS)

Mottram, R.; Millard, J.W.F.; Purdy, P.

2011-01-01

Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

Safety benefits from CANDU reactor replacement. A case study

International Nuclear Information System (INIS)

Mottram, R.; Millard, J.W.F.; Purdy, P.

2011-01-01

Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

Possibility of using gamma radiation from HTR reactors for the processing of food and medical products

International Nuclear Information System (INIS)

Pahladsingh, R.R.

2004-01-01

During the fission process in most of the presently operating nuclear reactors nuclear energy is converted into thermal energy and transferred to common steam cycles for power generation. As part of the fission process also α, β and neutrons particles are released from the nucleus; the release of gamma-rays is also a part of the fission process. In present nuclear reactors α, β, neutrons particles and particularly gamma-rays are not gainfully used as a result of the reactor design and of the containment. These plants are built as required by regulations and international standards for safety. The inherently safe HTR reactor, by its physics and design, does not need a special reinforced containment and it is worth looking into the possibilities of this design feature to use the by-products, such as Gamma-rays, from nuclear fission. In the HTR Pebble Bed Reactors the α, and β particles will remain in the kernels of the pebbles. This means that only the neutron particles and gamma-rays will be available outside the reactor pressure vessel. In this report a proposal is presented to use the gamma-rays of the HTR reactor for irradiation of food and agricultural produce. For neutron shielding a reflector is placed inside the reactor while outside the reactor neutron- and thermal-shielding will be accomplished with water. The high energy gamma-rays will pass through the water-shield and could be harnessed for radiation processing of food and medical products. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

International Nuclear Information System (INIS)

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1993-March 31, 1994

International Nuclear Information System (INIS)

1994-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

International Nuclear Information System (INIS)

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

Energy Technology Data Exchange (ETDEWEB)

Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

1995-03-01

The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

Science.gov (United States)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control

Status report on nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-01-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

1977-05-01

Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Radiation protection/shield design

International Nuclear Information System (INIS)

Disney, R.K.

1977-01-01

Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.) [de

Space-reactor electric systems: subsystem technology assessment

International Nuclear Information System (INIS)

Anderson, R.V.; Bost, D.; Determan, W.R.

1983-01-01

This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified

Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

2003-12-01

Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

Transfer hood for handling fuel assemblies in nuclear reactors and especially fast reactors

International Nuclear Information System (INIS)

Aubert, M.; Merland, D.; Renaux, C.

1975-01-01

A description is given of a hood for transferring fuel assemblies between two or more separate guide ramps inclined to the vertical at the same angle of slope and extending respectively through a first passage into the reactor vessel and through a second passage into a fuel-assembly storage chamber. The hood comprises at least one shielded duct joined to a flanged rotating portion which is so arranged that the open lower end of the shielded duct is positioned in the line of extension of one guide ramp and then the other as a result of displacement of the rotating portion

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

Micro-Reactor Physics of MOX-Fueled Core

International Nuclear Information System (INIS)

Takeda, T.

2001-01-01

Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

Preliminary design of reactor coolant pump canned motor for AC600

International Nuclear Information System (INIS)

Deng Shaowen

1998-01-01

The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

Training and research on the nuclear reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

1998-01-01

The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

Nuclear reactor system study for NASA/JPL

Science.gov (United States)

Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

1982-01-01

Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

Application of Bondarenko formalism to fusion reactors

International Nuclear Information System (INIS)

Soran, P.D.; Dudziak, D.J.

1975-01-01

The Bondarenko formalism used to account for resonance self-shielding effects (temperature and composition) in a Reference Theta-Pinch Reactor is reviewed. A material of interest in the RTPR blanket is 93 Nb, which exhibits a large number of capture resonance in the energy region below 800 keV. Although Nb constitutes a small volume fraction of the blanket, its presence significantly affects the nucleonic properties of the RTPR blanket. The effects of self-shielding in 93 Nb on blanket parameters such as breeding ratio, total afterheat, radioactivity, magnet-coil heating and total energy depositions have been studied. Resonance self-shielding of 93 Nb, as compared to unshielded cross sections, will increase tritium breeding by approximately 7 percent in the RTPR blanket and will decrease blanket radioactivity, total recoverable energy, and magnet-coil heating. Temperature effects change these parameters by less than 2 percent. The method is not restricted to the RTPR, as a single set of Bondarenko f-factors is suitable for application to a variety of fusion reactor designs

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Decommissioning of the Northrop TRIGA reactor

International Nuclear Information System (INIS)

Cozens, George B.; Woo, Harry; Benveniste, Jack; Candall, Walter E.; Adams-Chalmers, Jeanne

1986-01-01

An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

Directory of Open Access Journals (Sweden)

Košťál Michal

2016-01-01

Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

CRNL research reactor retrofit Emergency Filtration System

International Nuclear Information System (INIS)

Philippi, H.M.

1990-01-01

This paper presents a brief history of NRX and NRU research reactor effluent air treatment systems before describing the selection and design of an appropriate retrofit Emergency Filtration System (EFS) to serve these reactors and the future MX-10 isotope production reactor. The conceptual design of the EFS began in 1984. A standby concrete shielding filter-adsorber system, sized to serve the reactor with the largest exhaust flow, was selected. The standby system, bypassed under normal operating conditions, is equipped with normal exhaust stream shutoff and diversion valves to be activated manually when an emergency is anticipated, or automatically when emergency levels of gamma radiation are detected in the exhaust stream. The first phase of the EFS installation, that is the construction of the EFS and the connection of NRU to the system, was completed in 1987. The second phase of construction, which includes the connection of NRX and provisions for the future connection of MX-10, is to be completed in 1990

Fast reactor physics at CEA: present studies and future prospects

International Nuclear Information System (INIS)

Hammer, P.

1980-09-01

This paper aims at giving a general survey of the fast reactor core physics and shielding studies wich are in progress at CEA (1979-1983) in order to solve the neutronic problems related to: - core design optimization, - reactor operation and fuel management, - safety, for the development of fast commercial breeders in France after the SUPER-PHENIX 1 construction is achieved

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

Results of the Nucifer reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

2016-07-01

Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

Electric power from near-term fusion reactors

International Nuclear Information System (INIS)

Longhurst, G.R.; Deis, G.A.; Miller, L.G.

1981-01-01

This paper examines requirements and possbilities of electric power production on near-term fusion reactors using low temperature cycle technology similar to that used in some geothermal power systems. Requirements include the need for a working fluid with suitable thermodynamics properties and which is free of oxygen and hydrogen to facilitate tritium management. Thermal storage will also be required due to the short system thermal time constants on near-time reactors. It is possbile to use the FED shield in a binary power cycle, and results are presented of thermodynamic analyses of this system

Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

Energy Technology Data Exchange (ETDEWEB)

Caro, R

1976-07-01

In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

Shielding plug for LMFBR type reactors

International Nuclear Information System (INIS)

Hashiguchi, Ko.

1979-01-01

Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

Sources of gamma radiation in a reactor core

Energy Technology Data Exchange (ETDEWEB)

Roos, Matts

1959-05-15

In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

International Nuclear Information System (INIS)

Kyllingstad, G.

1977-01-01

The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

International Nuclear Information System (INIS)

John Darrell Bess

2008-01-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

Production of an economic high-density concrete for shielding megavoltage radiotherapy rooms and nuclear reactors

International Nuclear Information System (INIS)

Mortazavi, S. M. J.; Mosleh-Shirazi, M. A.; Maheri, M. R.; Haji-pour, A.; Yousefnia, H.; Zolghadri, S.

2007-01-01

In megavoltage radiotherapy rooms, ordinary concrete is usually used due to its low construction costs, although higher density concrete are sometimes used, as well. The use of high-density concrete decreases the required thickness of the concrete barrier; hence, its disadvantage is its high cost. In a nuclear reactor, neutron radiation is the most difficult to shield. A method for production of economic high-density concrete witt, appropriate engineering properties would be very useful. Materials and Methods: Galena (Pb S) mineral was used to produce of a high-density concrete. Galena can be found in many parts of Iran. Two types of concrete mixes were produced. The water-to-concrete (w/c) ratios of the reference and galena concrete mixes were 0.53 and 0.25, respectively. To measure the gamma radiation attenuation of Galena concrete samples, they were exposed to a narrow beam of gamma rays emitted from a cobalt-60 therapy unit. Results: The Galena mineral used in this study had a density of 7400 kg/m 3 . The concrete samples had a density of 4800 kg/m 3 . The measured half value layer thickness of the Galena concrete samples for cobalt 60 gamma rays was much less than that of ordinary concrete (2.6 cm compared to 6.0 cm). Furthermore, the galena concrete samples had significantly higher compressive strength (500 kg/cm 2 compared to 300 kg/cm 2 ). Conclusion: The Galena concrete samples made in our laboratories had showed good shielding/engineering properties in comparison with all samples made by using high-density materials other than depleted uranium. Based on the preliminary results, Galena concrete is maybe a suitable option where high-density concrete is required in megavoltage radiotherapy rooms as well as nuclear reactors

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

International Nuclear Information System (INIS)

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Remote maintenance considerations for swimming pool tokamak reactor

International Nuclear Information System (INIS)

Niikura, S.; Yamada, M.; Kasai, M.

1983-01-01

Swimming Pool Tokamak Reactor (SPTR) is one of the candidate devices which are expected to demonstrate physical and engineering feasibility for fusion power reactors. In SPTR, water shield is adopted instead of solid shield structures. Among the advantages of SPTR are, from viewpoint of remote maintenance, small handling weight and high space availability between TF coils and a vacuum vessel. On the other hand, high dose rate during reactor repair and adverse effects on remote maintenance equipment by the shielding water might be the disadvantage of SPTR, where it is assumed that the shielding water is drained during reactor repair. Since the design of SPTR is still at the preliminary stage, for remote maintenance, much effort has been directed to clarification of design conditions such as environment and handling weight. As for the remote maintenance system concepts, studies have been focussed on those for a vacuum vessel and its internal structure (blanket, divertor and protection walls) expected to be repaired more frequently. The vacuum vessel assembly is divided into 21 sectors and number of TF coils is 14. A pair of TF coils are connected with each other by antitorque beams on the whole side surface. Vacuum vessel cassettes and associated blanket, divertor and protection walls are replaced through seven windows between TF coils pairs. Therefore each vacuum vessel cassette is required moving mechanisms in toroidal and radial directions. Options for slide mechanisms are wheels, balls, rollers and water bearings. Options for driving the cassette are self-driving by hydraulic motors and external driving by rack-pinion, wires or specific vehicles. As a result of studies, the moving mechanism with wheels and hydraulic motors has been selected for the reference design, and the system with water bearings and rack-pinion as an alternative. Furthermore typical concepts have been obtained for remote maintenance equipment such as wall-mounted manipulators, tools for

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

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.

Flow and pressure profiles for the primary heat transport system of Rajasthan Atomic Power Station for the operation with few isolated reactor channels near the end shield cracks

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A J; Chaki, S K; Sehgal, R L; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

The RAPS (Rajasthan Atomic Power Station) unit-1 is now operating at reduced power due to the removal of fifteen fuel channels for repair of south end shield cracks. The power level is restricted to 50% of the full power capacity as a precautionary measure. The relative difference that operation at 50% power and higher power would make to the end shield structure is being currently analysed with a view to operate this reactor at higher power levels. As a prerequisite, a detailed thermal hydraulic analysis is essential to assess the effect of reactor operation with isolated channels on the primary heat transport (PHT) system pressure, flow, temperature. The adequacy of the existing trip set points for the plant operation under this mode is also required to be assessed. In the present study, analysis of the PHT system has been carried out to determine the flow and pressure profiles for the RAPS heat transport system for operation of the reactor with isolated channels. (author). 5 refs., 1 fig., 1 tab.

1DB, a one-dimensional diffusion code for nuclear reactor analysis

International Nuclear Information System (INIS)

Little, W.W. Jr.

1991-09-01

1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs

Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5

International Nuclear Information System (INIS)

Ninkovic, M.

1976-01-01

In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation [sr

Reference reactor module for NASA's lunar surface fission power system

International Nuclear Information System (INIS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

2009-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

Proceedings of the specialists' meeting on reactor group constants

Energy Technology Data Exchange (ETDEWEB)

Katakura, Jun-ichi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

This report is the Proceedings of the Specialists' Meeting on Reactor Group Constants. The meeting was held on February 22-23, 2001 at Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of 59 specialists. The evaluation work for JENDL-3.3 is going on for the publication in a short time. The processing JENDL-3.3 file to make reactor group constants is needed when it is used in application fields. In the meeting, the present status of the reactor group constants was reviewed and the issues relating to them were discussed in such fields as thermal reactor, criticality safety, fast reactor, high energy region, burn-up calculation and radiation shielding. At the final session in the meeting, standardization of reactor group constants was discussed and the need of the reference group constants was confirmed by the participants. The 11 of the presented papers are indexed individually. (J.P.N.)

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

Energy Technology Data Exchange (ETDEWEB)

McKenzie, J L [Armed Forces Radiobiology Research Institute, Bethesda, MD (United States)

1974-07-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the {sup 41}Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

International Nuclear Information System (INIS)

McKenzie, J.L.

1974-01-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the 41 Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Barranco R, F.

2015-01-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

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

1963-02-15

RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

Conceptual design study of fusion experimental reactor (FY86 FER)

International Nuclear Information System (INIS)

Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

1987-09-01

This report describes the results of the reactor configuration/structure design for the fusion experimental reactor (FER) performed in FY 1986. The design was intended to meet the physical and engineering mission of the next step device which was decided by the subcommittee on the next step device of the nuclear fusion council. The objectives of the design study in FY 1986 are to advance and optimize the design concept of the last year because the recommendation of the subcommittee was basically the same as the design philosophy of the last year. Six candidate reactor configurations which correspond to options C ∼ D presented by the subcommittee were extensively examined. Consequently, ACS reactor (Advanced Option-C with Single Null Divertor) was selected as the reference configuration from viewpoints of technical risks and cost performance. Regarding the reactor structure, the following items were investigated intensively: minimization of reactor size, protection of first wall against plasma disruption, simplification of shield structure, reactor configuration which enables optimum arrangement of poloidal field coils. (author)

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

2017-04-01

Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

Spring unit especially intended for a nuclear reactor core

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, Wilhelm.

1977-01-01

This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

Compact-toroid fusion reactor based on the field-reversed theta pinch: reactor scaling and optimization for CTOR

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1980-01-01

Early scoping studies based on approximate, analytic models have been extended on the basis of a dynamic plasma model and an overall systems approach to examine a Compact Toroid (CT) reactor embodiment that uses a Field-Reversed Theta Pinch as a plasma source. The field-reversed plasmoid would be formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high-technology plasmoid source from the hostile reactor environment. Stabilization of the field-reversed plasmoid would be provided by a passive conduction shell located outside the high-temperature blanket but within the low-field superconducting magnets and associated radiation shielding. On the basis of this batch-burn but thermally steady-state approach a reactor concept emerges with a length below approx. 40 m that generates 300 to 400 MWe of net electrical power with a recirculating power fraction less than 0.15

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

International Nuclear Information System (INIS)

Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

2010-01-01

LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

Management of the radioactive waste resulting from the Romanian VVR-S research reactor decommissioning

International Nuclear Information System (INIS)

Ene, D.; Cepraga, D.G.

2002-01-01

The paper consists in a waste study of the Romanian VVR-S reactor which will be prepared for decommissioning operations after the permanent shutdown (23.12.1997). Calculations were carried out to determine the activity arising from neutron activation of structural materials inside the reactor, considering the design of the facility and its operating rules. To this end, the following method was used: i) Neutron flux distribution within the reactor was calculated using the DORT transport code, based on DLC23 shielding library relating to three cylindrical reference systems of the reactor structure: reactor core, horizontal tube and thermal column; ii) Calculation of the activity of each reactor component at different cooling times was performed by the ANITA2000 code, using the neutron flux, compositional data for each material and the power history of the reactor; iii) Unconditional clearance indexes for all material at various cooling times were calculated using the clearance levels defined in IAEA-TECDOC-855; iv) Total activities and masses by material type, within the waste category and for each decay time were calculated by summation of the data previously classified for each reactor component. The resulting activation inventory and waste masses, falling in IAEA defined waste categories are presented in the paper at periods of 100 days, and 6, 10, 25, and 50 years after reactor the shutdown. For some components of the reactor as: aluminum central vessel, the central iron shielding ring, the time behaviour of both the fin spatial activity distribution and the radionuclide contributions to the total activity are plotted in the paper. (author)

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

1996-01-01

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

Reversed field pinch reactor study 3

International Nuclear Information System (INIS)

Hollis, A.A.; Mitchell, J.T.D.

1977-12-01

This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

2002-03-01

In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

Pailla, Henri; Schaller, Karl; Vidard, Michel.

1974-01-01

A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

Neutron activation of building materials used in the reactor shield

International Nuclear Information System (INIS)

Hernandez, A.T.; Perez, G.; D'Alessandro, K.

1993-01-01

Cuban concretes and their main components (mineral aggregates and cement) were investigated through long-lived activation products induced by neutrons from a reactor. The multielemental content in the materials studied was obtained by neutron activation analysis in an IBR-2 reactor and gamma activation analysis in an MT-25 microtron from Join Institute of Nuclear Research of Dubna. After irradiation of building materials for 30 years by a neutron flow of unitary density, induced radioactivity was calculated according to experimental data. The comparative evaluation of different concretes aggregates and two types of cement related to the activation properties is discussed

Investigation of small and modular-sized fast reactor

International Nuclear Information System (INIS)

Kubota, Kenichi; Kawasaki, Nobuchika; Umetsu, Yoichiro; Akatsu, Minoru; Kasai, Shigeo; Konomura, Mamoru; Ichimiya, Masakazu

2000-06-01

In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

Repairing method and apparatus for weld portion of reactor core shroud

Energy Technology Data Exchange (ETDEWEB)

Tsujimura, Hiroshi; Tamai, Yasukata; Kurosawa, Koichi (Hitachi Ltd., Tokyo (Japan)); Toyota, Seiichi; Kikuchi, Toshikazu.

1993-12-07

A method of repairing a weld portion in a cylindrical reactor core shroud comprises a first step of inspecting a weld portion by an ultrasonic flow testing device from the surface of the reactor core shroud, a second step of applying repairing fabrication for cracked portion if it is discovered by the test and a third step of applying a surface modification to the fabricated portion after the repairing fabrication. As a result, repairing fabrication for the crack caused by stress corrosion crack or the like is enabled and reoccurrence of the stress corrosion crack in the repair fabrication portion can be prevented. Operator's exposure dose is minimized by shielding with reactor water or shielding plate. In a case of using the shielding plate, welding and surface improvement can be practiced in atmospheric air instead of water-submerged welding. Water does not intrude from the outside of the shroud and occurrence of penetration crack can be coped with. Further, it is possible to reduce cost and save labors for parts exchange by using the parts in common, to improve the operation efficiency. (N.H.).

Repairing method and apparatus for weld portion of reactor core shroud

International Nuclear Information System (INIS)

Tsujimura, Hiroshi; Tamai, Yasukata; Kurosawa, Koichi; Toyota, Seiichi; Kikuchi, Toshikazu.

1993-01-01

A method of repairing a weld portion in a cylindrical reactor core shroud comprises a first step of inspecting a weld portion by an ultrasonic flow testing device from the surface of the reactor core shroud, a second step of applying repairing fabrication for cracked portion if it is discovered by the test and a third step of applying a surface modification to the fabricated portion after the repairing fabrication. As a result, repairing fabrication for the crack caused by stress corrosion crack or the like is enabled and reoccurrence of the stress corrosion crack in the repair fabrication portion can be prevented. Operator's exposure dose is minimized by shielding with reactor water or shielding plate. In a case of using the shielding plate, welding and surface improvement can be practiced in atmospheric air instead of water-submerged welding. Water does not intrude from the outside of the shroud and occurrence of penetration crack can be coped with. Further, it is possible to reduce cost and save labors for parts exchange by using the parts in common, to improve the operation efficiency. (N.H.)

Heat transfer in inertial confinement fusion reactor systems