International Nuclear Information System (INIS)

The performance of anaerobic biological process is heavily process conditions dependent. In this study, an attempt has been made to investigate the influence of process conditions like temperature, sludge age and hydraulic retention time (HRT) on the efficiency of an upflow anaerobic sludge blanket (UASB) reactor and upflow anaerobic sludge filter (UASF) to treat combined industrial wastewater. Reactors were operated at easing ambient temperatures (38, 30, 20 and 14 deg. C) and correspondingly increasing sludge ages (60, 90, 120 and 150 days). At temperature 38 deg. C and sludge age of 60 days, UASF showed better performance than VASE reactor. This mainly due to the enhanced filtration through well-graded sand filter and fairly good biological activity in UASF. At this stage, lack of sludge granulation in VASE reactor resulted in poor biological activity; hence, ...

Science.gov (United States)

Both ammonium and nitrite act as substrates as well as potential inhibitors of anoxic ammonium-oxidizing (Anammox) bacteria. To satisfy demand of substrates for Anammox bacteria and to prevent substrate inhibition simultaneously; two strategies, namely high or low substrate concentration, were carefully compared in the operation of two Anammox upflow anaerobic sludge blanket (UASB) reactors fed with different substrate concentrations. The reactor working at relatively low influent substrate concentration (NO(2)(-)-N, 240 mg-NL(-1)) was shown to avoid the inhibition caused by nitrite and free ammonia. Using the strategy of low substrate concentration, a record super high volumetric nitrogen removal rate of 45.24 kg-Nm(-3) day(-1) was noted after the operation of 230 days. To our knowledge, such a high value has not been reported previously. The evidence from transmission electron ...

Science.gov (United States)

In order to lower the costs for second generation bioethanol from lignocellulosic biomass anaerobic digestion of the effluent from ethanol fermentation was implemented using an upflow anaerobic sludge blanket (UASB) reactor system in a pilot-scale biorefinery plant. Both thermophilic (53 degrees C) and mesophilic (38 degrees C) operation of the UASB reactor was investigated. At an OLR of 3.5 kg-VS/(m(3) day) a methane yield of 340 L/kg-VS was achieved for thermophilic operation (53 degrees C) while 270 L/kg-VS was obtained under mesophilic conditions (38 degrees C). For loading rates higher than 5 kg-VS/(m(3) day) the methane yields were, however, higher under mesophilic conditions compared to thermophilic conditions. The conversion of dissolved organic matter (VS(diss)) was between 68% and 91%. The effluent from the ethanol fermentation showed no signs of ...

International Nuclear Information System (INIS)

This study describes the feasibility of anaerobic treatment of complex phenolics mixture from a simulated synthetic coal wastewater using four identical 13.5 L (effective volume) bench scale hybrid up-flow anaerobic sludge blanket (HUASB) (combining UASB + anaerobic filter) reactors at four different hydraulic retention times (HRT) under mesophilic (27 #+-# 5 "oC) conditions. Synthetic coal wastewater with an average chemical oxygen demand (COD) of 2240 mg/L and phenolics concentration of 752 mg/L was used as substrate. The phenolics contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. The study demonstrated that at optimum HRT, 24 h, and phenolic loading rate of 0.75 g COD/(m"3-d), the phenolics and COD removal efficiency of the reactors were 96% and 86%, respectively. Bio-kinetic models were applied to ...

Science.gov (United States)

The anaerobic transformation and degradation of nitrophenols by granular sludge was investigated in upflow anaerobic sludge blanket (UASB) reactors continuously fed with a volatile fatty acid (VFA) mixture as the primary substrate. During the start-up, subtoxic concentrations of 2-nitrophenol (2-NP), 4-nitrophenol (4-NP), and 2, 4-dinitrophenol (2, 4-DNP) were utilized. 4-NP and 2, 4-DNP were readily converted to the corresponding aromatic amine; whereas 2-NP was converted to nonaromatic products via intermediate formation of 2-aminophenol (2-AP). These conversions led to a dramatic detoxification of the mononitrophenols because the reactors treated the nitrophenolics at the concentrations which were over 25 times higher than those that caused severe inhibition. VFA removal efficiencies greater than 99% were achieved in both reactors at loading rates greater than 11.4 g COD per liter of ...

Science.gov (United States)

The complexity and diversity of the microbial communities in biogranules from an upflow anaerobic sludge blanket (UASB) bioreactor were determined in response to short-term changes in substrate feeds. The reactor was fed simulated brewery wastewater (SBWW) (70% ethanol, 15% acetate, 15% propionate) for 1.5 months (phase 1), acetate / sulfate for 2 months (phase 2), acetate-alone for 3 months (phase 3), and then a return to SBWW for 2 months (phase 4). Performance of the reactor remained relatively stable throughout the experiment as shown by COD removal and gas production. 16S rDNA, methanogen-associated mcrA and sulfate reducer-associated dsrAB genes were PCR amplified, then cloned and sequenced. Sequence analysis of 16S clone libraries showed a relatively simple community composed mainly of the methanogenic Archaea (Methanobacterium and Methanosaeta), members of the Green Non-Sulfur (Chloroflexi) group of Bacteria, ...

Energy Technology Data Exchange (ETDEWEB)

Information is presented concerning reactor research activities; isotope geology; NERC radiocarbon laboratory; teaching activities; and reactor operation.

Energy Technology Data Exchange (ETDEWEB)

The costs of establishing a computer-based training program for FFTF reactor operators are analyzed.

Energy Technology Data Exchange (ETDEWEB)

This is a presentation on Human Factors in reactor operations. It discusses issues that deal with power plant operations, training and design, operational effectiveness and safety, supporting people to achieve effective and error free performance.

International Nuclear Information System (INIS)

The invention concerns a procedure for operating reactors in nuclear power plants. It aims at utilizing power reserves in nuclear power plants. This can be achieved by a steam-side connection of the steam generators of two reactors. The amount of steam exchanged between the units is chosen in such a way that power changes at the steam turbines feedback mainly to the corresponding reactor. In order to realize a high power transfer it is necessary to return the amount of condensate produced in the steam receiving unit and corresponding to the power transferred to the feedwater system of the steam donating unit.

Science.gov (United States)

Milestones concerning equipment, reactor components, and testing and operations at the FFTF since July 1, 1975 are highlighted. (JWR)

Energy Technology Data Exchange (ETDEWEB)

This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

International Nuclear Information System (INIS)

The document presented defines the plan to be used to coordinate the preparation, review, approval, and issuance of the operating procedure documents required to ensure safe and efficient operation of FFTF.

International Nuclear Information System (INIS)

The TVR heavy water research reactor was deployed at Moscow Institute of Theoretical and Experimental Physics. In 1990, the final batch of the spent nuclear fuel from this reactor was shipped to Production Association (PA) 'Mayak' for reprocessing. The SNF removal was a stage of the reactor decommissioning activities. The designs of the TVR reactor and its fuel elements are similar to the RA reactor designs. Two ways of the RA reactor SNF transportation to PA 'Mayak' have been considered: in aluminum barrels and in additional canisters using respectively TUK-32 and TUK-19 shipping casks. The practical experience and the equipment used to prepare for the TVR reactor SNF removal can be helpful to the RA reactor personnel in finding the best way to perform these engineering operations. (author)

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year ...

International Nuclear Information System (INIS)

Various schemes of cooling have been investigated for the purpose of assessing potential benefits on the operational characteristics of the Syrian MNSR reactor. A detailed thermal hydraulic model for the analysis of MNSR has been developed. The analysis shows that an auxiliary cooling system, installed in the pool which surrounds the lower section of the reactor vessel, will significantly offset the consumption of excess reactivity due to the negative reactivity temperature coefficient, Hence, the maximum operating time of the reactor is extended. Compared with experimental data, the suggested model proves to be valid for the analysis of MNSR behavior under both steady state and transient conditions. (author)

Energy Technology Data Exchange (ETDEWEB)

Results of tests on the 1 kg/h continuous reactor for the hydroliquefaction of coal are described. The reactor was operated at 415-435 C and 21 MPa using a continuous stirred reactor with a retention time of about 2 hours. All product oils were recovered by distillation. Sub-bituminous coal was found to give the best product yield. Tests using 5% red mud and 3% improved red mud showed significant increases in oil yield. (4 refs.)

Energy Technology Data Exchange (ETDEWEB)

An experimental plan for improving the problems of failed fuel location system in Wolsung Unit-2 reactors was established. It is not possible to make an experiment on the failed fuel monitoring nuclides in the cold laboratories because they have very short half life. Therefore, the experiments can be only carried out at the existing monitoring system under reactor operation. For that reason, an experimental plan was drawn up for installing the radiation detection system on reactor site.

International Nuclear Information System (INIS)

Since the first round of conceptual fusion reactor designs in 1973 - 1974, there has been considerable progress in design improvement. Two recent tokamak designs of the Wisconsin and Culham groups, with increased plasma beta and wall loading (power density), lead to more compact reactors with easier maintenance. The Reference Theta-Pinch Reactor has undergone considerable upgrading in the design of the first wall insulator and blanket. In addition, a conceptual homopolar energy storage and transfer system has been designed. In the case of the mirror reactor, there are design changes toward improved modular construction and ease of handling, as well as improved direct converters. Conceptual designs of toroidal-multiple-mirror, liner-compression, and reverse-field pinch reactors are also discussed. A design is presented of a toroidal multiple-mirror reactor that ...

Energy Technology Data Exchange (ETDEWEB)

The operation of a nuclear power reactor necessarily implies the consumption or burnup of reactor fuel by fission and capture, which gives rise to a decrease in the reactivity of the reactor. The effect of americium formation on the criticality of a thermal power reactor using two types of fuel is studied. The three-dimensional core calculation is used to calculate the production of the transuranium isotopes and their effect on the effective multiplication factor (K[sub eff]). This effect cannot be neglected for thermal power reactors with UO[sub 2]-PuO[sub 2] fuel (3.11% after 70 weeks of operation). The effect of the transuranium isotopes on the K[sub eff] for a thermal power reactor with UO[sub 2] fuel is about 0.0018% and can be ignored. (author).

British Library Electronic Table of Contents (United Kingdom)

In allusion to the explosion of a urea reactor took place in a fertilizer plant at Pingyin, Shandong, China, a series of evidence collection and inspection jobs which includes collecting operation condition and parameters, sampling the explosion fracture, reactor body apart from explosion fracture, and leak detection medium and its hangover, etc., had been carried out firstly. Based on these jobs, farther analysis and computation work has been done to the structural and materials characteristics and the operation condition of the urea reactor, including compositions, metallographic phases, tensile properties, impact energy, strain ageing characteristics, and fracture toughness of the urea reactor steels, the compositions of leak detection medium and its hangover in the urea reactor, and ex...

International Nuclear Information System (INIS)

Criticality was first achieved with the Fast Flux Test Facility (FFTF) a little more than 10 yr ago on February 9, 1980. Although the FFTF was designed and built primarily for testing fuels, materials, and components for the liquid-metal fast breeder reactor program, it has, over its first 10 yr of operation, provided valuable information in many other areas. This paper provides a summary of the contributions to the physics of liquid-metal reactors (LMRs) obtained from operation of and testing in the FFTF, with emphasis on some of the more significant and interesting accomplishments.

International Nuclear Information System (INIS)

Dr. Rudolf Schulten of West Germany and expert witnesses from national laboratories, utilities, and the nuclear industry testified on reactor safety issues as they relate to the Chernobyl accident and public concern that modern technology has not paid enough attention to public safety. Each of the witnesses contributed safety-related information based on what has been learned from the Soviet incident. Particular focus went to similarities and differences between the Chernobyl and US reactors in safety design and engineering and to the environmental effects of the accident. The N reactor near Richland and a commercial reactor at Fort St. Vrain, Colorado are the only two operating graphite reactors, but neither is a boiling water reactor.

Science.gov (United States)

The Guide is intended to provide guidelines for the initial preparation of FFTF Operating Procedures. The Procedures Preparation Guide was developed from the plan presented and approved in the FFTF Reactor Plant Procedures Plan, PC-1, Revision 3.

Energy Technology Data Exchange (ETDEWEB)

This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

Energy Technology Data Exchange (ETDEWEB)

The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in ...

Energy Technology Data Exchange (ETDEWEB)

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

British Library Electronic Table of Contents (United Kingdom)

A new chemical reactor, the Open Plate Reactor, is being developed by Alfa Laval AB. It combines good mixing with high heat transfer capacity into one operation. With the new concept, highly exothermic reactions can be produced using more concentrated reactants. A nonlinear model of the reactor is derived and a control system is developed. For temperature control a cooling system is designed and experimentally verified, which uses a mid-ranging control structure to increase the operating range of the hydraulic equipment. A Model Predictive Controller is proposed to maximize the conversion under hard input and state constraints. An extended Kalman filter is designed to estimate unmeasured concentrations and parameters. Simulations show that the designed control system gives high conversion ...

International Nuclear Information System (INIS)

Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986).

Energy Technology Data Exchange (ETDEWEB)

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.

Energy Technology Data Exchange (ETDEWEB)

The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test ...

Energy Technology Data Exchange (ETDEWEB)

In this paper the possibility of configuring a water cooled Nuclear Thermal Propulsion (NTP) rocket, based on a Particle Bed Reactor (PBR) is investigated. This rocket will be used to operate on water obtained from near earth objects. The conclusions reached in this paper indicate that it is possible to configure a PBR based NTP rocket to operate on water and meet the mission requirements envisioned for it. No insurmountable technology issues have been identified.

International Nuclear Information System (INIS)

A brief review of apparatuses used at enterprises engaged in industrial processing of spent nuclear fuel for dissolving dioxide nuclear fuel from power reactors is provided. Advantages and drawbacks of facilities operating in periodic, semi-continuous and continuous modes are considered. It is pointed out that today there are two promising trends in developments in the field, i.e. rotor- and vibrational-type dissolving apparatuses operated continuously

Energy Technology Data Exchange (ETDEWEB)

Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described.

Energy Technology Data Exchange (ETDEWEB)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

International Nuclear Information System (INIS)

Nuclear physics parameters of the Wolsung CANDU-PHW reactor are computed by use of the PHWCELL computer code that is an improved version of LATREP. The PHWCELL code mainly computes cell parameters of heavy water moderated reactors, and modeling scheme of heavy water reactor cell calculations has been developed with the PHWCELL computer code. The reactor operating conditions considered in the study are cold zero power (CZP) and hot full power (HFP) with equilibrium poison. The cell parameters are also computed as a function of fuel burnup and the numerical results are compared with the results in PSR of the Wolsung unit and in the previous study. (author).

International Nuclear Information System (INIS)

In this paper the diffusion of tritiated water vapor into concrete walls is studied to evaluate tritiated water retention capacity of a fusion reactor concrete building. Using a model of the tritiated water diffusion determined form experimental results, depth profiles of tritiated water in concrete are calculated in the case of being exposed to air containing tritiated water vapor during the normal operational condition of a fusion reactor. A 0.5-m-thick concrete is sufficient for reactor hall walls from a viewpoint of the tritium containment.

International Nuclear Information System (INIS)

This paper describes the Fast Flux Test Facility (FFTF) Operations initial training program utilizing the Operator Training Simulator (OTS). The OTS is a computer-driven system that provides real time response of essential FFTF plant functions to a control room mockup. The FFTF, a 400 Megawatt, three-loop, sodium-cooled fast test reactor will test fuels, materials and equipment for the U.S. Liquid Metal Fast Breeder Reactor Program. Construction is expected to be completed in August 1978. Initial criticality is expected in early 1979. This schedule will require FFTF control room operators to be fully qualified to operate the facility by late 1979. Because FFTF is like no other U.S. nuclear reactor, existing U.S. utility plants could not be depended on to provide highly experienced people to operate FFTF. Therefore, an ...

International Nuclear Information System (INIS)

Atomic Energy Regulatory Board carries out the regulatory review of the reactor physics design, commissioning and operational aspects through Project Design Safety Committee and Specialist Group of reactor physicists with wide experience in the design, commissioning and operational safety review of NPPs. TAPP-3 and 4 PHWRs, being the first indigenous design of 540 MWe Units, are quite different than the standard 220 MWe PHWRs. The safety review of reactor physics design was quite complex, as majority of the systems were new. The Reactor Physics Specialist Group carried out extensive safety review of 540 MWe PHWR reactor physics design and made significant contributions of design modifications and improvements in the operational procedures. Some salient contributions include: Monitoring the core during bulk addition of ...

Science.gov (United States)

The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the passive emergency core cooling system. The decay heat is removed by the conventional active low-pressure residual heat removal system. As for a postulated severe accident, the suppression pool water floods the lower part of the reactor pressure vessel (RPV) in the case when core damage occurs, and the in-vessel retention that keeps the melt inside the RPV is achieved by supplying the coolant. The containment adopts a parallel-double-steel-plate structure similar to a hull ...

International Nuclear Information System (INIS)

Argentina has two operating PHWR nuclear power plants. Atucha I NPP is a pressure vessel type heavy water reactor of 360 MW e with 25 years of operation and Embalse NPP is a pressure tube type CANDU-600 reactor of 640 MW e. Atucha II, a third plant of 600 MW e of the pressure vessel type similar to Atucha I, is being constructed. NASA (Nucleoelectrica Argentina S.A.) currently operates both nuclear power plants. The National Atomic Energy Commission (Comision Nacional de Energia Atomica - CNEA) provides operational support to the plants, including research and development assistance, and actual technical services and maintenance work in different areas. The Chemistry Department, formerly the Reactor Chemistry Department has carried out project and support activities to the plants during the past 20 years. The aim of this work is to describe ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station unit-4 is first 540 MWe pressurized heavy water reactor. Heavy water is used as the coolant and moderator. With reactor operation tritium is formed by absorption of neutron by deuterium atoms. Experience in the 220 MWe indicates that tritium is main contributor to the internal dose of radiation workers. Study on tritium build up in Primary Heat Transport (PHT) and Moderator (MOD) system was carried out at the initial stage of the operation of the unit-4. This paper brings out tritium activity buildup in the PHT and MOD systems and its comparison with 220 MWe reactors. This paper helps in estimation of the internal dose contribution to the radiation workers at TAPS 3 and 4. (author)

Energy Technology Data Exchange (ETDEWEB)

The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design ...

Energy Technology Data Exchange (ETDEWEB)

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be ...

Energy Technology Data Exchange (ETDEWEB)

The deliberation by the Nuclear Safety Commission was initiated on the alteration in reactor installation, as was required by the Ministry of International Trade and Industry. The alteration is the additional installation of the reactor No. 2 in the Sendai Nuclear Power Station, Kyushu Electric Power Co., Inc. It is a PWR power plant with thermal output of about 2,660 MW (electric output of 890 MW), to be installed, adjoining to the reactor No. 1 of the same type and capacity under construction. In the examination by MITI, it was confirmed that the technological capabilities for its construction and operation and the radiation protection measures in power generation are both sufficient. The contents of the examination include the siting conditions, the location and construction of reactor facilities, etc.

Energy Technology Data Exchange (ETDEWEB)

The renovation programme of the Phenix nuclear power plant (fast neutrons reactor situated at Marcoule) has for objective to ensure the reactor operation lengthening. In this frame, expertise and monitoring operation in situ of materials have been started. The presence of sodium and a temperature at the cold breakdown of the primary circuit between 150 and 180 degrees (Celsius) imply, for fast reactors, very special conditions. In this context, Framatome has realised three intervening in the area of nondestructive testing: the inspection of the cone-shaped support ring, the monitoring of the upper part of the primary vessel and the monitoring of the intermediary exchanger equipment. (N.C.)

Science.gov (United States)

This request is submitted to seek interim approval to operate a Toxic Substances Control Act (TSCA) of 1976 chemical waste landfill for the disposal of polychlorinated biphenyl (PCB) waste. Operation of a chemical waste landfill for disposal of PCB waste ...

International Nuclear Information System (INIS)

Apppendix B of 10 CFR Part 50 establishes quality assurance requirements for the operation of nuclear power plant safety-related structures, systems and components. This Guide describes an acceptable method for complying with these regulations with regard to overall quality assurance program requirements for the operation phase of nuclear power plants. Input to this Guide has been provided by the Advisory Committee on Reactor Safeguards.

International Nuclear Information System (INIS)

The activation aspects of pure fusion and hybrid fusion technology is studied to assess the radioactive safety of various fusion concepts including tokamak pure fusion, fissile fuel producing hybrid and radio waste transmuting hybrid. The activation properties of breeding, coolant and structural materials in fusion reactors might be quite different from those in fission reactors because of the high energy D-T fusion neutrons from the fusion cores. A study on the involved activation reactions and the uncertainties of the associated nuclear cross-sections is carried. The activation properties of various first wall concepts and blanket concepts are discussed. The radioactive inventory during the operation lifetime and the potential hazard of the radioactive nuclides with respect to near term (reprocessing) and long term (waste disposal) aspects are calculated, with reference to ITER/NET (International Thermonuclear Experiment ...

International Nuclear Information System (INIS)

Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are ...

Energy Technology Data Exchange (ETDEWEB)

The report presents a summary of the work of the Department of Nuclear Safety Research and Nuclear Facilities in 1995. The department`s research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au) 5 tabs., 21 ills.

British Library Electronic Table of Contents (United Kingdom)

Main results of the 40-year experience gained from operation of horizontal steam generators in VVER-type reactor installations used in Russia and many foreign countries are described. Existing unresolved problems are pointed out.

Energy Technology Data Exchange (ETDEWEB)

This paper describes the advent of energy pumped storage plants in France, in connection with nuclear power reactors development and gives detailed specifications of different components (pump turbines, alternators, valves ...). 5 figs., 3 tabs.

British Library Electronic Table of Contents (United Kingdom)

The main problems encountered during the operation of horizontal steam generators are considered. Design features of the new PGV-1000MK and PGV-1500 steam generators are analyzed.

Energy Technology Data Exchange (ETDEWEB)

The WBURN (2-D, 2-group, coarse mesh) code is developed to analyze the equilibrium core characteristics of CANDU-PHWR. The equilibrium characteristics of Wolsung reactor computed by using WBURN are compared with the values given in the Wolsung FSR. The changes of equilibrium core characteristics caused by the variation of design parameters for operating conditions are also investigated. The numerical results indicate that the average discharge irradiation in the Wolsung reactor can be increased up to about 5%.

International Nuclear Information System (INIS)

This paper reports on a Level 1 PRA performed on the Omega West Reactor at Los Alamos National Laboratory. A Master Logic Diagram was used to identify possible initiating events. A chi-square distribution was used to quantify initiating event frequencies given that no initiating events have occurred in 30 years of OWR operation. The PRA results are presented as both probability density function and cumulative distribution function curves.

International Nuclear Information System (INIS)

The use of nuclear reactors to provide electrical energy has shown considerable growth since the first nuclear plant started commercial operation in the mid 1950s. Although the main purpose of this paper is to review the fuel cycle capabilities in the United States, the introduction is a brief review of the types of nuclear reactors in use and the world-wide nuclear capacity.

International Nuclear Information System (INIS)

A report on the Canadian Nuclear Society Conference on robotics and remote handling in the nuclear industry, September 1984. Remote handling in reactor operations, particularly in the Candu reactors is discussed, and the costs and benefits of use of remote handling equipment are considered. Steam generator inspection and repair is an area in which practical application of robotic technology has made a major advance. (U.K.).

Energy Technology Data Exchange (ETDEWEB)

The steam generator low water level trip setpoint of KORI NPP units 3 and 4(KNU 3 and 4), three-loop Westinghouse pressurized water reactor, is higher than that of OPR1000. In addition, steam generator downcomer water level in KNU 3 and 4 could fluctuate easily during a transient because of smaller downcomer water inventory, compared to the total water inventory in the steam generator. Due to these reasons, there is a higher possibility of unnecessary reactor trips caused by the steam generator low-low water level in KNU 3 and 4. Its operating history shows that most of reactor trips were caused by steam generator low-low level reactor trip signal. Such reactor trips, especially unnecessary ones, result in time and economic losses. In this paper, a feasibility study was performed to reduce unnecessary reactor trip by changing steam generator ...

International Nuclear Information System (INIS)

Generally, Hardening and irradiated brominating phenomena are occurred in the reactor vessel under operation conditions by atomic cavities and creation of impurity atoms which are led by high fast neutron flux. To assure the mechanical integrity of pressure vessel until the end of power plant life after monitoring the sample specimens on the vessel inside, a series of tests is performed over the retrieved surveillance capsule to examine the changes according to the plant operation in accordance with regulations. Monitoring surveillance capsules attached to neutron shield wall of outer core are consists of impact sample, tensile sample and temperature monitor

Energy Technology Data Exchange (ETDEWEB)

Shield measurements and four cycles of operating experience have shown the design and construction of radiation shields for the Fast Flux Test Facility (FFTF) reactor and plant to be satisfactory. A number of minor shield deficiencies were found and corrected. Most of these were associated with interfaces between components, each of which was satisfactory by itself. Preliminary evaluation of the shield measurements indicates satisfactory agreement with design calculations. Operator doses to date have been quite small, especially when compared to light water reactor experience.

Energy Technology Data Exchange (ETDEWEB)

In order to improve the well-known short operational life time of BF3 counters, three potential adsorbents for impurity gases (graphite, activated charcoal and a zirconium-aluminum mixture) were introduced into BF3 counters in the form of coating on the aluminum cathode surface. Tests in el fields revealed that a partial coating of activated charcoal provides the best result. The improvement of their operational life in el fields was about three orders of magnitude in terms of tolerable exposure. Many counters with a partial coating of activated charcoal were further tested from the following viewpoints: background noise, vibration and shock, el pulse discrimination, operational life in a neutron field and non-operational in-reactor exposure life. The results were satisfactory for reactor control and protection usage. (author).

International Nuclear Information System (INIS)

Full text: Research reactors have been playing a multi dimensional role in areas of nuclear fuel cycle programme, radio-isotope productions, neutron beam research etc. To ensure an efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are required on routine basis. 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 requires a prior estimation of the reactivity load due to the sample, heating rate and the activity developed in it during irradiation. For the safety of the personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be ...

International Nuclear Information System (INIS)

... power reactors pressure tube reactors reactors THERMAL REACTORS.

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

TAPP-3 and 4 reactors use large number of Self Powered Neutron Detectors (SPNDs) for Neutronic lower measurement and control. To perform in-situ calibration of these detectors in select locations and to validate the reactor physics codes which predict flux at various points in the core, traveling in-core probes (TIP) are required. The TIP assembly consists of a miniature neutron sensitive detector. The detector is driven in and out of core using a mechanism which facilitates positioning of the detector anywhere inside a vertical tube (Central carrier tube of any of the six select Vertical Flux Units) in the core. TIP is driven through retractable feed mechanism for a stroke of 13 m. This paper describes the developmental efforts and the operational feedback of the retractable feed mechanism for the stroke of 13 m used at TAPP 3 and 4 reactor. (author)

Energy Technology Data Exchange (ETDEWEB)

A comprehensive review was conducted of primary helium coolant chemistry data, based on current and past operating histories of helium-cooled, high-temperature reactors (HTGRs), including the Fort St. Vrain (FSV) HTGR. A reference observed FSV reactor coolant environment was identified. Further, a slightly drier expected FSV coolant chemistry was predicted for reactor operation at 100% of full power. The expected environment was compared with helium test environments used in the US, United Kingdom, Germany, France, and Japan. Based on a comprehensive review and analysis of mechanical property data reported for Alloy 800 tested in controlled-impurity helium environments (and in air when appropriate for comparison), an assessment was made of the effect of FSV expected helium chemistry on material properties of alloy 800, with emphasis on design properties of the Alloy 800 material ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400 MW(th) sodium cooled reactor and is the largest test reactor of its type in the world. It was designed and is being operated to serve two purposes: gaining liquid metal system experience and serving as a test bed for fuels and materials. During test operations it is possible that cladding breaches and escape of fission gas to the reactor cover gas region can occur. To identify the source of such a leak all 78 fuel pin assemblies contain ''gas tag'' with a unique ''tag'' mixture in each assembly. The mass spectrometric identification of tag isotope ratios makes possible rapid location and thus faster removal (if required) of breached test pins.

British Library Electronic Table of Contents (United Kingdom)

A dense Pd-Ag membrane reactor (MR) with 100% hydrogen selectivity packed with either Rh/La2O3 or Rh/La2O3-SiO2 as catalysts was used to carry out the dry reforming of methane. The membrane reactor simulation was performed using a well-known reactor model. For this purpose, we employed the equations derived from complete kinetic studies of the dry reforming of methane reaction in connection with both catalysts. In addition, we developed the kinetic equation for the reverse water gas shift reaction (RWGS). The combination of detailed kinetic studies with the measured permeation flux for the Pd-Ag membrane allowed a complete comparison between experimental and simulated operation variables. The variables studied for both catalysts were methane conversion and hydrogen permeation as a function...

Energy Technology Data Exchange (ETDEWEB)

Analyses were performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse and General Electric commercial reactors during the period 1984 through 1995. RPS operational data from these reactors were collected from the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LER). The common-cause failure (CCF) modeling in the fault trees developed for these studies and the analysis and use of common-cause failure data were sophisticated, state-of-the-art efforts. The overall CCF effort helped to test and expand the limits of the U.S. Nuclear Regulatory Commission's CCF methodology.

Energy Technology Data Exchange (ETDEWEB)

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering ...

International Nuclear Information System (INIS)

In General Electric's design of the Simplified Boiling Water Reactor (SBWR), The depressurization valves (DPV) are installed in the reactor pressure boundary: four are connected to the reactor vessel by means of nozzles, and two more are located on the main steam pipes (one DPV for each line), which act during particular transients and/or loss of coolant accidents (LOCA), consequently providing the reactor vessel with a safe quick depressurization system. Once the vessel is de pressurised, the passive gravity-driven cooling system (GDCS) starts to operate, permitting the injection of water required for continuous core cooling. DPVs are leak tight, with welded flaps, actuated by a [striker[hammer***] which is activated by an explosive mixture. The dynamic loads that open these valves include, in addition to those produced by steam (typical in any thermodynamic transient with ...

Energy Technology Data Exchange (ETDEWEB)

The PERMCAT process chosen for the final clean-up stage of the Tokamak Exhaust Processing system of the ITER tritium plant combines in a single component a catalytic reactor and a permeator using Pd/Ag membranes. This study covers the mechanical behaviour of a Pd/Ag membrane under different operating conditions. The consequences of hydrogen uptake by the membrane during nominal operation but also during off-normal events are presented. Depending on the operating conditions, expansions around 2% and significant deformations are observed. Different mechanical designs of PERMCAT reactors are then discussed. The first generation comprises finger-type membranes and two new mechanical designs use either additional edge welded bellows or a special corrugated Pd/Ag membrane. These upgraded designs improve the robustness and simplify the geometry of the component. The experimental validation ...

International Nuclear Information System (INIS)

Exposure control at the operating Nuclear Power Station is a major concern. TAPS Unit-4 is the first Pressurized Heavy Water Reactor of 540 MWe electrical capacity. This unit was made critical on March 6, 2005. In-depth review of radiation safety was done to identify the impact of design modification on dose rates at various locations and on the equipment's. Problems encountered in controlling the dose rates in 220 MWe electrical are eliminated by appropriate design modifications. Due to higher capacity of the unit there are design changes in major systems such as reactor core, primer heat transport system, moderator system, reactor regulation and protection systems. Reactor operations and maintenance activities during shut down contributes to exposure of the employees. Based on the experience gained in the operation of 220 MWe, design ...

International Nuclear Information System (INIS)

Korea Atomic Energy Research Institute(KAERI) launched a project to develop an integral reactor in 1996. The reactor called as System Integrated Modular Advanced Reactor(SMART) which is a kind of small modular reactors (SMRs). Since the early 1990s, there has been renewed interest in the development and application of small and medium sized integral reactors. 2009 assessment by the IAEA under its Innovative Nuclear Power Reactor and Fuel Cycle (INPRO) program concluded that there could be 96 SMRs in operation around the world by 2030 in its 'high' case, and 43 units in the 'low' case, none of them in the USA. The reason of the increased demand mostly comes from the fact that SMRs are thought to be more suitable for developing countries with small electrical grid capacity, insufficient infrastructure and limited investment capability than ...

International Nuclear Information System (INIS)

A LOss of Electric Power(LOEP) experiment was conducted after a 30MW full power operation as one of the reactor performance tests to verify the design characteristics of the HANARO. The objective of LOEP test was to investigate the integral behaviors of the system and the components as well as the cooling characteristics when the electric power was lost unexpectedly. Through the test, it was confirmed that the residual heat from the core was safely removed by the natural convection cooling and the assistant power systems operated normally

Science.gov (United States)

This request is submitted to seek interim approval to operate a chemical waste landfill for the disposal of polychlorinated biphenyl (PCB) wastes. This request covers only the disposal of small quantities of solid PCB wastes contained in decommissioned su...

International Nuclear Information System (INIS)

The sequence of events from the assembly of a space nuclear power system to its integration in the Space Shuttle Transportation System (STS) is considered. First, the sequence followed for SNAP-10A, the only free world space reactor electric power system ever launched and operated in space, is reviewed. Before shipment, the SNAP-10A reactor was raised to operating temperature using electrically supplied heat and operated at low power for control calibration. Next we discuss shipment to the launch site, a phase that is critical because of the potential for various accidents. Once the power system arrives at the launch site, the processing sequence is performed. This sequence includes checkout, mating with the payload or upper stage launch vehicle, and integration into the STS.

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400 Mwt sodium-cooled fast reactor operating at the Hanford Engineering Development Laboratory, Richland, Washington, to conduct fuels and materials testing in support of the US Liquid Metal Fast Breeder Reactor (LMFBR) program. Startup and initial power testing included a comprehensive series of nonnuclear and nuclear tests to verify the thermal, hydraulic, and neutronic characteristics of the plant. A specially designed series of natural circulation tests were then performed to demonstrate the inherent safety features of the plant. Early in 1982, the FFTF began its first 100-day irradiation cycle. Since that time the plant has operated very well, achieving a cycle capacity factor of 94% in the most recent irradiation cycle. Seventy-five specific test assemblies and 25,000 individual fuel pins have been irradiated, some in excess of 80 MWd/Kg.

Energy Technology Data Exchange (ETDEWEB)

The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and ...

Energy Technology Data Exchange (ETDEWEB)

The Fixed Bed Nuclear Reactor (FBNR) is a small 40 MWe reactor based on the Pressurized Water Reactor (PWR) technology. FBNR is an integrated primary circuit and simple in design. It has the characteristics of being small, modular, proliferation resistant, inherently safe and passively cooled reactor with reduced adverse environmental impact. It utilizes the fuel designed for high temperature reactors operating in a relatively low temperature of PWR environment The 15 mm diameter spherical fuel elements are made of TRISO type microspheres embedded in graphite and cladded by SiC. The coolant flow transfers them from the fuel chamber into the core and become fixed forming a suspended core. Any accident signal will cut off the power to the coolant pump causing a stop in the flow. This results in making the fuel elements fall out of the reactor ...

International Nuclear Information System (INIS)

GRR-1 is a 5MW open pool type research reactor with MTR-type fuel elements cooled and moderated by light water with beryllium reflectors at the two opposing sides of the core. A graphite thermal neutron column is adjusted to one side of the core. Six radial horizontal beam tubes are available, of which three contain in-pile collimators for neutron scattering instruments. The reactor is currently out of operation for inspection and refurbishment purposes. The core has been dismantled and the fuel elements are stored in the used fuel storage tank. The GRR-1 inspection and refurbishment plan involves inspection and eventually replacement of the reactor's primary cooling circuit. The health physics procedures to be implemented during inspection of the main water outlet are divided in three stages: a) pool dose rate survey from pool top, b) pool drainage by decreasing water level in steps and c) inspection ...

International Nuclear Information System (INIS)

In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR"T"M or ATMEA"T"M designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR"T"M reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR"T"M can be ...

International Nuclear Information System (INIS)

The second international conference on containment design and operation included sessions on the following topics: performance and regulatory requirements; radionuclide behaviour; severe accident design and analysis; operation, maintenance, leaking and aging of containment systems; thermal hydraulic behaviour of containment systems; hydrogen mixing and mitigation; design methods and concepts; code validation; structural analysis and response tests; passive safety systems; aerosol behaviour; containment reliability, integrity, and risk assessment; hydrogen deflagration and detonation. Due prominence was given to CANDU and other PHWR reactors. The individual papers have been abstracted separately.

Energy Technology Data Exchange (ETDEWEB)

The objective of Configuration Risk Management Program(CRMP) is to maintain the safety level by assuring the defense-in-depth of nuclear power plant while the configurations are changed during plant operations, especially for the LP/SD. Such a safety purpose can be achieved by establishing the risk monitoring programs with both quantitative and qualitative features. Generally, the quantitative risk evaluation models, i.e., PRA models are used for the risk evaluation during full power operation, and the qualitative risk evaluation models such as safety function assessment trees are used. Through this study, safety function assessment trees were developed.

Energy Technology Data Exchange (ETDEWEB)

Optimum plant operation can often be achieved by means of constraint control instead of model- based on-line optimization. This is because optimum operation is seldom at the top of the hill but usually at the intersection of constraints. This article describes the development of a constraint control system for a plant producing ethylbenzene (EB) by the Mobil/Badger Ethylbenzene Process. Plant optimization can be defined as the maximization of a profit function describing the economics of the plant. This function contains terms with product values, feedstock prices and operational costs. Maximization of the profit function can be obtained by varying relevant degrees of freedom in the plant, such as a column operating pressure or a reactor temperature. These degrees of freedom can be varied within the available operating margins of the plant.

International Nuclear Information System (INIS)

In the future even more than in the past, nuclear power will be indispensable in the present industrialized countries and in those still under development. The safe, nonpolluting, and economic supply of energy to mankind in the future includes so many different problems that the technology of the high-temperature reactor at its present level of development, and with the possibilities is offers above and beyond those provided by other, established, technologies, does not have to mark the end of some old line of development, but rather should be seen as the starting point of a development offering promise for the future. It is for this very reason that the extensive, valuable knowledge and experience accumulated in the construction, operation, and decommissioning of the AVR and THTR plants, the development of the HTR module and other variants and, last, but not least, the valuable results of projects such as PNP, NFE, and HHT, must be preserved ...

International Nuclear Information System (INIS)

The Swedish State Power Board has together with Nukem, Hanau, West-Germany carried out pyrolysis o powder resins in a pilot plant with a capacity of about 30 kg/hr. The pyrolysis reactor with its afterburner and offgas scrubber system has been operated under steady state condition. About 2200 kg resins have been pyrolysed under November-December 1983 and the decontamination factor for Cs has been measured. Solidification of the residues from the pyrolysis reactor and scrubber water solutions has been carried out and various recipes with cement have been tested. The pyrolysis process has high decontamination factors and no offgas problems as the operating temperature of the reactor is Low. The residues from the reactor are chemically dead and can not cause swelling problems. Compared with a normal cementation process the final waste volume will be reduced with a ...

Energy Technology Data Exchange (ETDEWEB)

The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design ...

International Nuclear Information System (INIS)

The study of the dynamics of lubricants and mechanical components encased in metal enclosures is important to many industries. Of particular importance is the flow characteristics of oils or similar lubricants within the metal enclosure during operation of the device. The purpose of this summary is to report on the design and successful application of a real-time neutron radiography system to study the lubrication and design of the piston and seal of a gas spring. In addition, the application of this technique to a wider range of similar problems using the pulse capability of the TRIGA reactor is described.

British Library Electronic Table of Contents (United Kingdom)

The search for the mixing angle Formula Not Shown , the last unknown angle in the neutrino mixing matrix, is one of the main priorities in the field of neutrino physics. By measuring Formula Not Shown to better than 0.01 at 90% C.L., the Daya Bay Reactor Neutrino Experiment has the highest sensitivity to this parameter among all the other experiments that are currently operating or under construction. The experiment consists of multiple identical detectors placed underground at different baselines from three groups of reactors, a configuration that minimizes systematic errors and cosmogenic backgrounds. The main aspects of the experiment, as well as its current status and future prospects, are reviewed.

International Nuclear Information System (INIS)

Mixed Uranium Plutonium Carbide ((U, Pu) C), in the form of pellets encapsulated in stainless steel tubes is the fuel for Fast Breeder Test Reactor (FBTR) at Kalpakkam. For the fabrication of fuel for enlarging the core of this reactor, high burn up plutonium is used. The external exposure in these labs was significantly higher than that with low burn up Pu fuel. Dose evaluation to the organs was carried out using experimental TLDs during various operations of FBTR fuel fabrication to study the dose distribution pattern. (author)

Science.gov (United States)

This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate.

Energy Technology Data Exchange (ETDEWEB)

The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through February 1999.

International Nuclear Information System (INIS)

The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs.

Energy Technology Data Exchange (ETDEWEB)

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the Halden Reactor Project. In 1958, the first Halden Reactor Project Agreement was signed by organisations representing 12 European countries. During 1994 France became a full member and associate membership was established with Russia. Accordingly, 16 countries were participating in the Project by the end of the year. The objectives have evolved from being simply a demonstration of the operation of a boiling heavy-water reactor to becoming a substantial research and development programme covering the domains of human-machine interaction, fuel behaviour, materials testing, water chemistry, and instrumentation. In 1994, significant progress was achieved in all of the areas addressed by the project, including the re-instrumentation of irradiated fuel rods, fission gas ...

Energy Technology Data Exchange (ETDEWEB)

This technology development plan is designed to provide a clear understanding of the research and development direction necessary for the qualification of nuclear grade graphite for use within the Next Generation Nuclear Plant (NGNP) reactor. The NGNP will be a helium gas cooled Very High Temperature Reactor (VHTR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Considerable effort will be required to ensure that the graphite performance is not compromised during operation. Based upon the perceived requirements the major data needs are outlined and justified from the perspective of reactor design, reatcor performance, or the reactor safety case. The path forward for technology development can then be easily determined for each data need. How the data will be obtained and the inter-relationships between the experimental ...

International Nuclear Information System (INIS)

Application of Programmable Digital Comparator System (PDCS) in NPP is to monitor large number of plant parameters and generate contact outputs for reactor trip, reactor setback, process interlocks etc. when parameters cross their operational bounds. Till NAPS these functions are achieved through individual Indicating Alarm Meters (IAM). PDCS used for the first time in KAPS replaces these IAMs. Since its inception, PDCS has undergone improvements in design, incorporates additional functionalities/enhanced features. Dedicated PDCS is provided in TAPP-3 and 4 for protection function. System re-configurability, on-line inter channel comparison of safety critical process parameters' values, Graphic User Interface etc. are other enhancements. From KAPS to TAPP-4 system has given many years of almost trouble-free operation. Commissioning of TAPP-3 system has been very smooth. Objective of the paper is to ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF)is a 400 MW (th) sodium-cooled fast flux test reactor located on the Hanford Site in southeastern Washington State. The FFTF is operated by the Westinghouse Hanford Company for the United States Department of Energy. The FFTF is a three loop plant designed primarily for the purpose of testing full-scale core components in an environment prototypic of future liquid metal reactors. The plant design emphasizes features to enhance this test capability, especially in the area of the core, reactor vessel, and refueling system. Eight special test positions are provided in the vessel head to permit contact instrumented experiments to be installed and irradiated. These test positions effectively divide the core into three sectors. Each sector requires its own In-Vessel Handling Machine (IVHM) to access all the core positions. Since the core and the in-vessel refueling components ...

Energy Technology Data Exchange (ETDEWEB)

One of the issues identified in the US Nuclear Regulatory Commission`s Nuclear Plant Aging Research program plan is the need to understand the state of ``mothballed`` or other out-of-service equipment to ensure subsequent safe operation. Programs for proper storage and preservation of materials and components are required by NRC regulations (10 CFR 50, Appendix B). However, materials and components have been seriously degraded due to improper storage, protection, or layup, at facilities under construction as well as those with operating licenses. Pacific Northwest Laboratory has evaluated management of aging for unstarted or mothballed nuclear power plants. The investigations revealed that no uniform guidance in the industry addresses reactor layup. In each case investigated, layup was not initiated in a timely manner, primarily because of schedule uncertainty. Hence, it is reasonable to assume that this delay resulted in ...

Energy Technology Data Exchange (ETDEWEB)

The AVR experimental nuclear reactor was Professor Dr. Rudolf Schulten's brainchild. Visionary ideas led to the success of this technology: - Helium coolant because of the particularly high heat transfer coefficients; - an integrated primary system reactor concept as the basis of all safety considerations in the interest of maximum safety; - uranium-235 and thorium-232 fuel allowing new fuel to be bred; - high temperatures for electricity generation at maximum thermodynamic efficiencies, i.e. optimum fuel utilization; - the possibility to run chemical processes economically at high temperatures by means of nuclear fuels; - the inherent safety of the reactor, for a major accident accompanied by a complete loss of cooling cannot occur for nuclear physics reasons, as was tested twice in the AVR. The AVR attained its first criticality on August 28, 1968. It was operated for more than 20 years, ...

International Nuclear Information System (INIS)

Full text: The nuclear renaissance, anticipated by AREVA in the beginning of the century is now happening in several countries around the world. The fundamentals being the increasing demand of energy, the volatility of fossil fuel prices, the awareness of climate change threat connected with the extensive use of fossil fuels. The EPRTM reactor present significant improvements compared to previous generation reactors enabling to reach an outstanding safety level (redundancy of safety systems, airplane crash resistance), to improve the economics (extended plant lifetime, flexibility and availability during operation and, increased efficiency and fuel utilization) while limiting the impact on workers and the environment. Several countries have been implementing the transition to third generation reactors. The presentation will analyze different examples in order to draw the lessons learned from this first ...

International Nuclear Information System (INIS)

Microprocessor based ''smart'' pressure, level, and flow transmitters were tested to determine the radiation hardness of this class of electronic instrumentation for use in reactor building applications. Commercial grade Complementary Metal Oxide Semiconductor (CMOS) integrated circuits used in these transmitters were found to fail at total gamma dose levels between 2500 and 10,000 rad. This results in an unacceptably short lifetime in many reactor building radiation environments. Radiation hardened integrated circuits can, in general, provide satisfactory service life for normal reactor operations when not restricted to the extremely low power budget imposed by standard 4--20 mA two-wire instrument loops. The design of these circuits will require attention to vendor radiation hardness specifications, dose rates, process control with respect to radiation hardness factors, and non-volatile programmable ...

International Nuclear Information System (INIS)

The IGC Highlights briefly outlines some of the the significant progresses made by Indira Gandhi Centre for Atomic Research, Kalpakkam during the period 1996-1997. The Fast Breeder Test Reactor (FBTR) was operated at the maximum power level possible with the available partial core. The first generation of electricity from FBTR and its synchronization with the grid in 1997 marked a significant step in the nuclear programme of the Centre. Another important event was the commissioning of the "2"3"3U - fuelled Kamini reactor.The mission-oriented programmes in fast reactor technology was supported by a host of research and development programmes, in closely related areas namely materials technology, welding metallurgy, sodium technology, manufacturing technology, non-destructive testing, quality engineering, in-service inspection, electronics and instrumentation and safety research. The Highlights also ...

Energy Technology Data Exchange (ETDEWEB)

The combustin of chloroethane is modeled as a stirred reactor so that we can study critical emission characteristics of the reactor as a function of residence time. We examine important operating conditions such as pressure, temperature, and equivalence ratio and their influence on destructive efficiency of chloroethane and production of other chlorinated products. The model uses a detailed chemical kinetic mechanism that we have developed previously for C{sub 3} hydrocarbons. We have added to this mechanism the chemical kinetic mechanism for C{sub 2} chlorinated hydrocarbons developed by Senkan and coworkers. Some reactions have been added to Senkan's mechanism and some of the reaction-rate expressions have been updated to reflect recent developments in the literature. In the modeling calculations, sensitivity coefficients are determined to find which reaction-rate constants have the largest effect on destructive ...

Energy Technology Data Exchange (ETDEWEB)

At present, two types of reactors, Pressurized Light Water Reactor(PLWR) and Pressurized Heavy Water Reactor(PHWR), are operating and under construction in Korea. They are much different in design concepts and inherent features from each other so that the calculation methods for Exclusion Area Boundary(EAB) are also different from each other. Thus, the domestic calculation methodology has been applied to PHWR, Wolsung 2, 3 and 4. In this report, the regulatory requirements and methodologies for EAB of Canadian methodology for EAB has been also investigated. It has been examined that the Canadian methodology which has been applied to the calculation of EAB of Wolsung 2, 3 and 4 can be said to be conservative enough compared to physical phenomena. 4 tabs., 3 figs., 22 refs. (Author).

CERN Document Server

The Daya Bay reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors worldwide and in reactors in Germany. Extensive laboratory testing and in-service experience suggest that the Alloy 800 tubing has excellent resistance to corrosion-related degradation under appropriate operating conditions. In planning refurbishment of nuclear plants stations, a key concern is the longevity of existing SGs up to the 60-year lifetime of the refurbished plant. The paper reviews an existing methodology based on the concept of the improvement factor, and refines its estimation based on data specific to CANDU operating conditions. The paper presents a more advanced Bayesian probabilistic approach to estimate the degradation free lifetime distribution of Alloy 800 tubing, which is used to quantify the probability of degradation during the service life and to evaluate the impact of ...

Energy Technology Data Exchange (ETDEWEB)

Fission energy systems that significantly reduce the need for the user country to be involved in the nuclear operations and technology could simplify implementation and reduce the proliferation potential. Conceptual system designs with improved (relative to the once-through LWR fuel cycle) proliferation resistance for application in developing countries are being evaluated. The fission energy systems being studied include all activities and equipment necessary to produce energy, recycle selected materials, and dispose of the waste. The systems currently being studied are required to function with no refueling of the reactors on the user site. These requirements are being used to initiate the study, on the assumption that removal of these operations from within the developing countries will improve the proliferation resistance. Preliminary evaluations of a small fast reactor core cooled either by sodium ...

Energy Technology Data Exchange (ETDEWEB)

Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors and in reactors in Germany. Extensive laboratory testing and in-service experience suggest that the Alloy 800 tubing has excellent resistance to corrosion-related degradation under specified and appropriate operating conditions. In planning refurbishment of CANDU stations, a key concern is the longevity of existing SGs up to the 60 year lifetime of the refurbished plant. The paper reviews an existing methodology based on the concept of the improvement factor, and estimates it based on experimental data specific to CANDU operating conditions. The paper presents a more advanced probabilistic approach to estimate the degradation free lifetime distribution of Alloy 800 tubing, which is used to quantify the probability of degradation during the service life and to evaluate the impact of potential ...

International Nuclear Information System (INIS)

... computer calculations fftf reactor nonlinear problems reactor accidents reactor

International Nuclear Information System (INIS)

Most of all research reactors are immerged in the deep water pool to be a ultimate heat sink. At the neighbor of the reactor, some radio-active matters, such as Na-24, Ar-41, Mg-27, Al-28 and etc, may be generated by the neutron irradiation. Those radio-active isotopes may rise up to the pool water surface through the natural convection flow, which can make the radioactivity in the reactor hall rise high enough to concern about the health of people working in the reactor hall. When the irradiation test facilities are loaded or unloaded during a normal operation, the highly radio-activated primary coolant may flow out through the irradiation test holes on the top of the reactor. This also may be a main hazard source to make the working environment of the reactor hall bad. Making a hot water layer 1.5 ? 2.0 m thick at the top of ...

Energy Technology Data Exchange (ETDEWEB)

The first Russian reprocessing facility, known as RT-1 (located at PA-Mayak in Ural region) was started on the radiochemical plant base in 1977. Nowadays RT-1 is the sole operating reprocessing plant in Russia. The main features of RT-1 is its broad spectrum of reprocessing spent nuclear fuels (SNF). The following spent fuel types are reprocessed: -) SNF from PWR reactors (WWER-440) and FB reactor (BN-600); -) SNF of transport ship reactors; -) production reactors SNF; -) research reactor spent nuclear fuel. The world-known technological processes are used at RT-1, but there are the following distinctive features. First, the universality of the three technological lines which allows not only the reprocessing of various SNF kinds, but also to implement the combined reprocessing of some types of them. Secondly, extraction of neptunium during SNF reprocessing which ...

Energy Technology Data Exchange (ETDEWEB)

The world`s population of research reactors is growing old. Many have been adapted to serve new purposes over their lives, from testing materials for nuclear power programmes and supporting neutron physics experiments, to colouring gemstones, doping silicon and generating medical isotopes. In the first article of this survey of research reactor issues, Wilfried Krull from GKSS in Germany describes the effects on a reactor of supporting these changes in application as ``design ageing`` . Managing this and other symptoms of ageing to extend plant life is a key task for operators, and Krull discusses the efforts being made internationally to handle them. Eventually, terminal decline of one vital component can determine when a reactor has to be shutdown for refurbishment. For BR2 in Belgium, it was the beryllium matrix. Edgar Koonen from SCK-CEN explains work being done to replace it ...

International Nuclear Information System (INIS)

The FFTF is a 400-MWt sodium-cooled fast neutron flux test reactor located on the US government-owned Hanford Site in southeastern Washington state. The reactor is operated for the US Department of Energy by the Westinghouse Hanford Company. Since FFTF started routine operation in 1982, the commercially fabricated driver fuel has performed flawlessly to well beyond the design goal peak burnup of 80 MWd/kgM. The core average discharge exposure is now some 60% beyond the original design expectations and attests to the ruggedness and reliability of the mixed oxide fuel system. In Cycle 9 sixteen long-life assemblies were installed to begin the irradiation of mixed oxides in the advanced low-swelling alloy HT-9 as the Core Demonstration Experiment (CDE). Operation of the plant from initial startup testing to ten cycles of operation has confirmed that the nuclear ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400-MW(t) sodium-cooled fast reactor operating at the Hanford Engineering Development Laboratory in Richland, Washington, to conduct fuels and materials testing in support of the US liquid-metal fast breeder reactor program. Startup and initial power testing included a comprehensive series of nuclear and nonnuclear tests to verify the thermal and neutronic characteristics of the plant and to demonstrate its inherent safety features. Extensive reactor core characterization measurements were completed to provide the neutron and gamma spectra, fission rates, and other physics data needed to design and evaluate tests irradiated in the FFTF. A specially designed series of natural-circulation tests was performed to demonstrate the inherent safety features of the plant. Early in 1982 the FFTF began its first 100-d irradiation cycle. Since that time the plant has ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents control room alarms and displays, and records and prints data and analysis reports. The fission gas monitor system makes extensive use of commercially available hardware and software, providing a reliable and easily maintained system. The design provides extensive automation of previous manual operations, reducing the need for operator training and minimizing the potential for operator error. The dual nature of the system allows one monitor to be taken out of service for periodic tests or maintenance without ...

British Library Electronic Table of Contents (United Kingdom)

For reliable operation and the optimization of production, industrial fermentation processes require appropriate tools for monitoring the process in real time. This work presents the structure and operation of a soft sensor for the on-line monitoring of biomass and product concentration during salinomycin and bacitracin fermentation in an industrial, 80-m^3 batch reactor; moreover it provides a tool for evaluation of batch production verified in industrial application. The process estimation algorithm consists of decoupled growth and product models, which ensures an unbiased convergence of the estimator and the robustness of the model. The production of secondary metabolites is described with a non-structured model upgraded with a variable forgetting factor that demonstrated a successful e...

Energy Technology Data Exchange (ETDEWEB)

The first stage of the continuous coal hydrogenation unit has been used to test a number of coals with different processing strategies. This work has shown that conversion increases with product recycle, however after the second pass the increase is small but operability of the reactor is considerably improved. A kinetic model for the aromatic saturation of the recycle solvent in the second stage has been developed and will be used in the selection of conditions for oil upgrading processes. New insights into the structural composition of coal derived materials have been made due to the refinement of chromatographic or solubility separation analyses into routine operations and the development of a new technique in NMR spectroscopy.

International Nuclear Information System (INIS)

The central feature of the Fast Flux Test Facility (FFTF) is the fast test reactor (FTR), which is a liquid-sodium-cooled fast reactor providing high fast-neutron flux for irradiation testing of fuels and materials. The FTR also provides a means to develop breeder reactor core components and to gain reactor systems operating experience for future liquid-metal fast breeder reactors (LMFBRs). In the FTR core, there are 82 incore positions (within rows 1 through 6) available for driver fuel assemblies and/or test assemblies. In addition, there are three safety rods and six control rods located in rows 3 and 5, respectively, in the three symmetric core sectors. The FFTF has been successfully and continuously operated for more than 11 reactor cycles. For the first 8 cycles, the core loadings were composed of the mixed-oxide ...

International Nuclear Information System (INIS)

The second regular inspection of No.2 plant in Sendai Nuclear Power Station was carried out from October 5, 1987 to January 8, 1988. the parallel operation was resumed on December 8, 1987, 65 days after the parallel off. The facilities as the object of inspection were the reactor proper, reactor cooling system, measurement and control system, fuel facilities, radiation control facilities, waste facilities, reactor containment installation, and emergency power generation system. On these facilities as the object of inspection, the appearance, disassembling, leak, function, performance and other inspections were carried out, and abnormality was not found at all. The works related to this regular inspection were accomplished within the range of allowable dose based on the relevant laws. The main reconstruction works carried out during the period of this regular inspection were the change of the degree of ...

International Nuclear Information System (INIS)

Reactor Regulating System (RRS) of TAPP-3 and 4 (540 MWe PHWR) addresses issues of elaborate Flux Tilt Control as applied to large Reactor Cores in addition to the traditional Bulk Power (Actual Power) Control. The control of Bulk and Zonal Power by RRS through the use of Zonal Control Compartments (ZCCs) has been successfully demonstrated in the Indian PHWRs for the first time. Features like automation in Demand Power Maneuvering, Manual Movement of Reactivity Devices through the Human Machine Interface (HMI) and the supervised withdrawal of Shut-off Rods during Auto Criticality are also included. Special algorithms to measure and control the individual Zone Power and Bulk Power also form part of RRS algorithms. This paper describes the salient features of RRS of TAPP-3 and 4 and the improvement carried out based on the feedback of past 1 year of operation of TAPP-4 at around 90 % FP. (author)

International Nuclear Information System (INIS)

First structural wall material (FSW) materials in inertial fusion energy (IFE) power reactors will be irradiated under typical repetition rates of 1-10 Hz, for an operation time as long as the total reactor lifetime. The main objective of the present work is to determine whether a continuous-pulsed (CP) approach can be an efficient method in modeling the pulsed activation process for operating conditions of FSW materials. The accuracy and practicability of this method was investigated both analytically and (for reaction/decay chains of two and three nuclides) by computational simulation. It was found that CP modeling is an accurate and practical method for calculating the neutron-activation of FSW materials. Its use is recommended instead of the equivalent steady-state method or the exact pulsed modeling. Moreover, the applicability of this method to components of an IFE power plant subject to ...

International Nuclear Information System (INIS)

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a ...

Energy Technology Data Exchange (ETDEWEB)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for ...

International Nuclear Information System (INIS)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for ...

International Nuclear Information System (INIS)

An x-ray fluorescence spectrometer was designed and fabricated which nondestructively determines the thickness of aluminum cladding at small suspected thin spots in the inner or outer surface of actinide reactor tubes. The analysis method is based on the difference in absorption of actinide L/sub #alpha#/ and L/sub #beta#/ fluorescent x-rays in passing through the cladding. Calibration plots of the logarithm of the L/sub #beta#//L/sub #alpha#/ x-ray intensity ratio versus cladding thickness are linear to at least 40 mils for U-Al, U_3O_8-Al, and PuO_2-Al substrates. Accuracy and precision of the experimentally determined cladding thickness and evaluated for both uranium and plutonium substrates. Experimental thickness data are reported for 618 quality assurance analyses on six Mark 41 PuO_2-Al target tubes. An x-ray fluorescence cladding thickness monitor operated with a computer-controlled fluoroscope holds considerable promise for quality ...

Energy Technology Data Exchange (ETDEWEB)

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides ...

Energy Technology Data Exchange (ETDEWEB)

A concept for a fast spectrum irradiation facility has been developed for insertion in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The design is based on the very large fast flux that is available in this reactor combined with the use of a strongly-absorbing thermal neutron shield. The preferred concept from the several considered consists of a three-pin design surrounded by a Eu{sub 2}O{sub 3} thermal neutron shield located in the reactor flux trap. Preliminary analyses showed that this concept can provide a fast flux larger than 1x10{sup 15} n/cm{sup 2}{center_dot}s and a fast-to-thermal flux ratio greater than 300 while having an acceptable impact on the HFIR operation. Additional analyses are necessary to confirm that this design is feasible and meets the requirements for fast fuel irradiation. If the design proves to be suitable, it can provide a relatively low-cost, near-term ...

Energy Technology Data Exchange (ETDEWEB)

Scenarios are developed for spent fuel that is taken out of the nuclear reactor and stored for 20 years before reprocessing, and three storage techniques which use a water pool, dry cask or vault are compared with respect to their costs. The storage price (storage cost per kilogram of spent fuel), which is employed as the economic index, is calculated on the assumption that all the charge is paid when the spent fuel is brought in the storage facilities. Four scenarios are assumed for spent PWR and BWR fuels to be stored in at-reactor (AT) or away-from-reactor (AFR) facilities. The capital costs cover the buildings of the storage facilities, equipment, decommissioning, casks for transportation and storage (for cask storage) and casks for transportation (for water pool or vault storage). Costs for operation and maintenance of the facilities are also considered in evaluating these methods. Evaluation ...

International Nuclear Information System (INIS)

The interrelations between the factors causing the main effects on the primary circuit coolant flow rate distribution in the horizontal steam generator pipes in reactor facilities with the WWER type reactors under the modes with natural circulation are discussed. The criterion showing the presence or absence of coolant circulation reversal in bottom rows of the steam generator pipes is obtained. It is shown that large hydraulic non-uniformity in steam generator pipes operating in parallel under coolant natural circulation leads to decreasing the heat transfer surface efficiency under reactor facility emergency cooling, restricts its servicing capabilities. The circulation reverse in steam generator pipes under coolant natural circulation mode can give unfavourable effect on separate structural elements of the steam generators and as a result it can cause additional temperature strains in metal. The ...

International Nuclear Information System (INIS)

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the ...

Energy Technology Data Exchange (ETDEWEB)

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

International Nuclear Information System (INIS)

The philosophy of containing tritium and activated products very close to the source and of operating by remote techniques is justified by a comparison with other concepts on protection and availability points of view. Several design options are studied according to the optimization protection methodology of ICRP. Provided that an important technological development is accomplished, the utilization of robotics and the limitation of containment volumes should be generalized.

International Nuclear Information System (INIS)

Appendix B to 10 CFR Part 50 establishes overall quality assurance requirements for the design, construction and operation of safety-related structures, systems, and components. This guide presents a method acceptable to the Commission for complying with these regulations with regard to overall quality assurance program requirements during design and construction of nuclear power plants. Input to this guide has been provided by the Advisory Committee on Reactor Safeguards.

International Nuclear Information System (INIS)

As the Operating and Support staffs of the FFTF organization have gained experience, the plant reliability and capacity factors have shown a steadily improving trend. The plant capacity factor for Cycle 4 was 99.5%. It is the purpose of this report to describe the evolution of the maintenance organization at the FFTF site from a general support organization to a technically proficient organization playing a major role in planning and performance of plant maintenance evolutions.

Energy Technology Data Exchange (ETDEWEB)

The mid-range industrial market currently consumes 4.2 million metric tons of hydrogen per year and has an annual growth rate of 15% industries in this range require between 100 and 1000 kilograms of hydrogen per day and comprise a wide range of operations such as food hydrogenation, electronic chip fabrication, metals processing and nuclear reactor chemistry modulation.

International Nuclear Information System (INIS)

In order to implement effective safety regulations on aging management for reactor facilities etc., the information on important technology issues, the latest technical knowledge including evaluation technology, test and research outcomes, related codes and standards, regulation information, operation experiences such as accidents and trouble, etc. with respect to aging-induced deterioration in and outside Japan and in other industries, were collected, organized and evaluated. (author)

Energy Technology Data Exchange (ETDEWEB)

The 1993-1995 period represents a time of significant transition for workers engages in nuclear-related activities. Total nuclear-related employment fell markedly from about 300,000 workers in 1993 to 265, 000 in 1995, a 12% decline. This is the lowest level since the 1970s. In 1995, a third of all nuclear-related employment was in the reactor operation and maintenance segment. 6 figs, 1 table.

Science.gov (United States)

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters.

Energy Technology Data Exchange (ETDEWEB)

The mission of the Heavy Water portion of D Area (or 400 Area) at SRS is to purify the site inventory of heavy water for storage in the Reactor Areas for future DOE missions.

Science.gov (United States)

The Savannah River Site (SRS) is a 310-square-mile United States Department of Energy nuclear facility located along the Savannah River (SRS) near Aiken, South Carolina. Nuclear weapons material production began in the early 1950s, utilizing five production reactors. In the early 1990s all SRS production reactor operations were terminated. The first reactor closure end state declaration was recently institutionalized in a Comprehensive Environmental Response and Compensation and Liability Act (CERCLA) Early Action Record of Decision. The decision for the final closure of the 318,000 square foot 105-P Reactor was determined to be in situ decommissioning (ISD). ISD is an acceptable and cost effective alternative to off-site disposal for the reactor building, which will allow for consolidation of remedial action wastes generated from other cleanup activities within ...

International Nuclear Information System (INIS)

Full text: The transmutation of nuclear waste to reduce the burden on a geological repository is a relevant topic within the Program of Nuclear Safety Research of the Research Centre Karlsruhe. Several studies have confirmed that a high efficiency of transmutation of actinides is reached in fast neutron spectrum reactor system. Therefore, an important effort is dedicated to the study of transmutation strategies with different fast reactors and their associated technologies. Moreover, in international contexts as Generation IV International Forum (GIF) and Sustainable Nuclear Energy Technology Platform (SNETP), fast reactors are considered in the frame of sustainable development of nuclear energy and reduction of waste. The systems that are currently under investigation, in the frame of the different fuel cycle scenarios, are liquid metal cooled and gas cooled fast reactors as well as Accelerator Driven ...

International Nuclear Information System (INIS)

The long-term support required from Canada in the operation and maintenance of overseas CANDU nuclear generating stations after commercial operation has commenced is described, with reference entirely to the KANUPP reactor. This includes: technical support to station staff to increase plant reliability and maintainability; assistance with plant improvements; procurement of spares and consumables; and assistance with training programmes. This technical support may be provided by a small number of Canadian staff actually resident at the power station; by short-term visits of Canadian specialists to site and by technical and procurement services provided from Canada. Examples of technical problems experienced are given, showing typical services required from Canada.

International Nuclear Information System (INIS)

A brief FFTF project description is provided which includes general plant siting information, general layout, plant design parameters, description of principal systems and components, and description of support facilities. The current status of the FFTF project is provided, including status of plant construction, overall status of the plant checkout and test program, status of operating authorization and plant operating procedures and personnel, and status of reactor core components and experiments. Specific information on the acceptance test program and early program results is discussed. The role of FFTF in the future breeder program is described, including its objectives for verification of plant system and components designs and operability and use as an irradiation test facility.

International Nuclear Information System (INIS)

Object: To permit a valve operation test to be periodically made during plant operation without causing variations in plant power by detecting flow control valve defect on the basis of a valve aperture alteration instruction. Structure: Step signals which are equal in absolute value and opposite in sign are coupled to the input side of flow controllers provided on the recycling loops of two or more recycling flow control systems. With these inputs the aperture of the flow control valve on one side is increased (or reduced) while the aperture of the valve on the other side is reduced (or increased). As a result, the recycling flow rate in the loop on one side is increased (or reduced) while that on the other side is reduced (or increased). Whether the valve is normally operating or not is confirmed by checking the recycling flow rate and valve aperture. (Nakamura, S.).

International Nuclear Information System (INIS)

Full text: Korea Atomic Energy Research Institute (KAERI) has many R and D facilities in operation. including HANARO research reactor, radioactive waste treatment facility (RWTF), post-irradiation examination facility (PIEF) and irradiated material test facility (IMEF). Recently. nation-wide interest is focused on the safety and security of major industrial facilities. Safe operation of nuclear facilities is imperative because of the consequence of public disaster by radiological release/contamination, in case of an accident. Recently, Ministry of Science and Technology (MOST) of the Korean government announced amendments of Atomic Energy laws to enforce requirements of the physical protection and radiological emergency. All provisions on nuclear safety regulation and radiation protection are entrusted to the Atomic Energy Act(AEA). The Act is enacted as the main law concerning the safety regulation of nuclear ...

International Nuclear Information System (INIS)

The procedure and operation-by-operation methods of the quality control of structural and fuel materials, mixed fuel pellets of UO_2-PuO_2, fuel element cans made of the AISI-316 steel and ready fuel elements are described as well as spacer wires (steel AISI-316), cases of fuel assemblies (FA) and completed FAs. The methods are used in manifacturing fuel elements and FAs for the FFTF reactor. The RDT standards that regulate the structure and functioning of the system of fuel element and FA production management are outlined. Destructive analytical methods characterized by sufficient accuracy but low productivity are noted to represent a considerable share of operations. Some specialized means of nondestructive testing are developed, such as the gauge to measure the total plutonium content in a fuel element, neutron radiography deVice and a laser gauge to measure the FA dimensions. The experience gained served as a basis for ...

Energy Technology Data Exchange (ETDEWEB)

The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E&E) and Chemistry & Material Sciences (C&MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E&E and C&MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. ...

International Nuclear Information System (INIS)

Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top "1"6N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University ...

International Nuclear Information System (INIS)

The Nuclear Assurance Corporation (NAC) owns and operates four NAC-1 truck casks. These casks are used to ship spent reactor fuel assemblies and radioactive reactor-core components. The casks have been loaded or unloaded at a total of fifteen nuclear facilities in the United States. In addition, NAC has used another large, overweight-truck cask to ship radioactive reactor core components from a reactor to a waste burial site. There are many individual differences in the cask handling facilities at each of the reactor stations, nuclear research facilities and the storage and burial sites serviced. Various types of auxiliary lifting and handling devices for on-site cask operations have been required. The quality assurance requirements for the equipment used in interfacing casks with nuclear power plant facilities have become more stringent. ...

Energy Technology Data Exchange (ETDEWEB)

A phase 2 study was initiated to investigate surfactant-assisted coal liquefaction, with the objective of quantifying the enhancement in liquid yields and product quality. This publication covers the first quarter of work. The major accomplishments were: the refurbishment of the high-pressure, high-temperature reactor autoclave, the completion of four coal liquefaction runs with Pittsburgh No. 8 coal, two each with and without sodium lignosulfonate surfactant, and the development of an analysis scheme for the product liquid filtrate and filter cake. Initial results at low reactor temperatures show that the addition of the surfactant produces an improvement in conversion yields and an increase in lighter boiling point fractions for the filtrate.

International Nuclear Information System (INIS)

For shut-down safety of the turbine generator (securing of auxiliary power operation after load shut-down and preventing the reaching of overspeed after load shut-down with disturbed turbine governing system) additional measures compared to those for superheated steam turbines are required for turbine generators in plants with pressurized water reactor (PWR) as well as those with boiling water reactor (BWR) . Equipment is described (e.g. overspeed govern or selecting connection, vacuum breaker, bypass valves, intercepting valves) which, depending on the own conditions of the individual turbine generator (e.g. run-up time, vacuum, enclosed energy), may be applied alone or in jointly. (orig.).

Energy Technology Data Exchange (ETDEWEB)

An experimental and modelling study was performed for a reverse flow catalytic converter attached to a natural gas/diesel dual fuel engine. The catalytic converter had a segmented ceramic monolith honeycomb substrate and a catalytic washcoat containing a predominantly palladium catalyst. A one-dimensional single channel model was used to simulate the operation of the converter. The kinetics of the CO and methane oxidation followed first-order behaviour. The activation energy for the oxidation of methane showed a change with temperature, dropping from a value of 129 to 35 kJ/mol at a temperature of 874 K. The reverse flow converter was able to achieve high reactor temperature under conditions of low inlet gas temperature, provided that the initial reactor temperature was sufficiently high. (author)

International Nuclear Information System (INIS)

Ppercase(femaxi-iv) is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of ppercase(femaxi-iv) is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of ppercase(femaxi-iv) was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in ppercase(femaxi-iv) and some results of applications on the experimental data. ((orig.)).

British Library Electronic Table of Contents (United Kingdom)

This study deals with the neutronics analysis of the current core configuration of a 3MW TRIGA Mark-II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Safety Analysis Report (SAR) values. The comprehensive neutronics code system SRAC was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Cross-section data library generated from JENDL-3.2 were used. The validation of the model against benchmark experimental results is presented. The SRA...

Energy Technology Data Exchange (ETDEWEB)

Mass transfer to a wall of a horizontal rectangular channel reactor was investigated by the limiting current technique for Reynolds numbers ranging from 200 to 32000. Overall mass transfer coefficients at various mass transfer surface angles were obtained while the reactor was operated under isothermal and non-isothermal conditions. Dimensionless correlations were developed for isothermal flows from 25 to 55{sup o}C and for non-isothermal flows with applied temperature differences up to 30{sup o}C. In the laminar flow range natural convection dominated, but under turbulent conditions combined natural and forced convection prevailed. Mass transfer was approximately doubled under optimum selection of channel surface rotation, temperature gradient and flow rate. (author)

Energy Technology Data Exchange (ETDEWEB)

This report discusses the UTR-10 reactor at Iowa State University which went critical on low enriched uranium (LEU) fuel on August 14, 1991. However, subsequent to the criticality experiments the fuel plates started to discolor. In addition, roll pins used to lift the fuel assemblies were discovered to be cracked. It was determined that these problems were due to chemical agents in the primary coolant water. The roll pins were replaced by solid stainless steel pins. The primary coolant was replaced and the reactor is currently in operation. Surveillance specimens will be used to monitor any possible future discoloration. The high enriched fuel (HEU) is being prepared for eventual shipment to a high enriched fuel receiving facility.

Energy Technology Data Exchange (ETDEWEB)

Femaxi-iv is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of femaxi-iv is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of femaxi-iv was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in femaxi-iv and some results of applications on the experimental data. ((orig.))

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The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

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The conceptual design for MCP to be installed in the integral reactor SMART was carried out. Canned motor pump was adopted in the conceptual design of MCP. Three-dimensional modeling was performed to visualize the conceptual design of the MCP and to check interferences between the parts. The theoretical design procedure for the impeller was developed. The procedures for the flow field and structural analysis of impeller was also developed to assess the design validity and to verify its structural integrity. A computer program to analyze the dynamic characteristics of the rotor shaft of MCP was developed. The rotational speed sensor was designed and its performance test was conducted to verify the possibility of operation. A prototypes of the canned motor was manufactured and tested to confirm the validity of the design concept. The MCP design concept was also investigated for fabricability by establishing the manufacturing procedures. 41 refs., ...

International Nuclear Information System (INIS)

A detailed assessment of the FFTF Primary Heat Transport System (PHTS) piping has led to the conclusion that the integrity of the piping is assured such that there is no realistic potential for a rupture. Nevertheless, consistent with the practice of showing design margins even for hypothetical events, a spectrum of postulated PHTS ruptures has been analyzed. The analyses showed that upstream of the reactor vessel inlet downcomer, rupture areas of any size including a double-ended rupture could be tolerated with no core coolant boiling. At the most limiting location, the reactor inlet nozzle, rupture areas of 75 in."2 and 55 in."2 could be tolerated for three-loop and two-loop operation, respectively. This paper will present the following: (1) the criterion with which consequences of postulated pipe ruptures are compared; (2) the general transient response of the FFTF to postulated ruptures; and (3) the acceptable rupture ...

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This compilation contains 41 Advisory Committee on Reactor Safeguards (ACRS) reports submitted to the Commission, Executive Director for Operations, or to the Office of Nuclear Regulatory Research, during calendar year 1991. It also includes a report to the Congress on the NRC Safety Research Program. All reports have been made available to the public through the NRC Public Document Room and the US Library of Congress. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and by chronological order within project name. Part 2 categorizes the reports by the most appropriate generic subject area and by chronological order within subject area.

International Nuclear Information System (INIS)

Large sized reactor such as 540 MWe Pressurised Heavy Water Reactor (PHWR) requires continuous in core monitoring of local flux in order to provide effective control and protection. About 198 self powered neutron detectors (SPNDs) of the straight individually replaceable type are distributed in the reactor core. For purposes of reactor regulation, 42 prompt responding cobalt SPNDs called zone control detectors (ZCDs) are housed in vertical flux units (VFUs) and these are uniformly distributed in 14 power zones. The in core detectors used for spatial control by ZCCs do not accurately represent average zone power as they sense the flux over a small volume. Flux mapping system (FMS) comprising of 102 vanadium SPNDs in 26 VFUs, provide accurate measure of neutron flux, even though they have slow response to change in neutron flux levels. For reactor protection system-1 (RPS-1), 36 ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled, fast neutron test reactor located on the Hanford Site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents control room alarms and displays, and records and prints data and analysis reports. The Fission Gas Monitor System (FGMS) integrates commercially available hardware and software, providing a reliable and easily maintained system. The design provides extensive automation of previous manual operations, reducing the need for operator training and minimizing the potential for operator error. The dual nature of the system allows either system A or B to be taken out of service for periodic tests or ...

International Nuclear Information System (INIS)

... fftf reactor mockup multiplication factors reactivity worths reactor cores reactor

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On March 28, 1979, the Pennsylvania Three Mile Island nuclear power plant Unit 2 (TMI-2) suffered a partial fuel-melt accident. During this accident, over 20,000 lb of molten fuel flowed through holes melted through the baffle plates and through the lower-core support assembly (LCSA). The molten fuel subsequently resolidified in the bottom of the reactor vessel. The lower-core support assembly of the TMI-2 reactor was not structurally damaged during the accident. In order to permit defueling of that region of the core, the LCSA was cut to permit access. A five-axis teleoperator was developed to deliver plasma arc cutting, rotary grinding and abrasive waterjet cutting of end effectors to the LCSA. Complex geometry sectioning was completed in a mock-up facility at chemistry and pressure conditions simulating those of the vessel, prior to actual in-vessel operations. In-vessel activities began in early May 1988 and were ...

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The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power Plants.`` The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available ...

International Nuclear Information System (INIS)

A commercial CFD code, Fluent, has been used to analyse the design of a filter-press reactor operating with characteristic linear flow velocities between 0.024 and 0.192 m s-1. Electrolyte flow through the reactor channel was numerically calculated using a finite volume approach to solve the Navier-Stokes equations. The length of the channel was divided into 7 sections corresponding to distances of 0, 0.01, 0.04, 0.08, 0.12, 0.14 and 0.15 m from the electrode edge nearest to the inlet. The depth of the channel was divided into three planes parallel to the channel bottom. For each channel section, a velocity profile was obtained at each depth together with the average velocity in each plane. The flow predictions show that the flow development, as the electrolyte passes through the cell, is strongly affected by the manifold causing strong vortex structures at the entrance and exit of the channel. Although the flow ...

International Nuclear Information System (INIS)

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety of the facility. The three passive safety attributes being evaluated in the paper are: 1) In-core and in-vessel natural convection cooling, 2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and 3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond to most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) models and results were ...