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Sample records for 500mwe fbr metal

  1. Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

    Research highlights: → ULOF analysis of metal (U-Pu-6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR. → Uncertainties (typically 20%) on the sensitive feedback parameters. → Sensitive parameters - core radial feedback and sodium void reactivity effect. → Transient behavior of both 500 MWe and 1000 MWe core are benign under ULOFA. → For 1000 MWe inherent safety is assured with limited sodium void reactivity. - Abstract: Unprotected loss of flow (ULOF) analysis of metal (U-Pu-6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.

  2. A comparative study of unprotected loss of flow accidents in 500 MWe FBR metal cores with PFBR oxide core

    Analyses of unprotected loss of flow accidents for 500 MWe U-Pu-6%Zr and U-Pu-10%Zr metal fueled sodium cooled reactors are presented and compared with that of the 500 MWe (U-Pu) MOX Prototype Fast Breeder Reactor (PFBR - under construction in Kalpakkam, India). A flow halving time of 8 s is considered for all the cases. It is found that the results of the metal fuel cases are close to each other. The loss of flow accident is benign for the metal fueled reactors where it is found that sodium coolant boiling is delayed up to 900 s, without credit for safety grade decay heat removal systems. In contrast, the oxide fueled reactor shows much earlier onset of sodium boiling and fuel slumping, leading to near prompt criticality and entry into the disassembly phase. Thus it is concluded that unprotected loss of flow accidents in metal fueled reactors are benign and allow sufficient time for operator action, if safety grade decay heat removal systems are able to remove the decay heat.

  3. Reactor protection systems of 500 MWe PHWRs

    The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author)

  4. Power distribution monitoring and control in 500 MWe PHWR

    The 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) is expected to be commissioned in a few years. It has a relatively large sized core with complex material distribution in comparison to the currently operating 220 MWe PHWRs. The resulting neutronically loosely coupled system demands continuous control of the core power distribution. This paper gives a brief description and analysis of the reactor monitoring and control system proposed for this reactor. (author). 11 refs, 8 figs, 3 tabs

  5. Inherent safety aspects of metal fuelled FBR

    Highlights: Inherent safety of metal fuelled FBR is studied by static and dynamic methodology of reactor physics and thermal-hydraulics. ► It is discovered that FBR with metal fuel is inherently safe against ULOFA. ► Sensitive parameters are core radial expansion feedback, sodium void effect and flow halving time. ► Sensitivity analyses are carried out with 20% uncertainty. ► Inherent safety of 1000 MWe with the extended flow coast down is recommended to avoid cliff edge effects. -- Abstract: Static and dynamic studies of metal fuelled fast breeder reactors (MFBR) are carried out to verify the passive shutdown capability and its inherent safety parameters. Static calculations are carried out to determine the vested reactivity feedback parameters from the fuel and coolant temperature rise separately. Power reactivity decrement of metal fuel reactor is found to be small as compared to oxide fuel reactor of same size. ULOF analysis of metal (U–Pu–6% Zr) 1000 MWe pool type MFBR is studied with a flow halving time of 8 s. The study is also made with considering uncertainties on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behaviour of 1000 MWe core is benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal System (SGDHRS) capacity before the initiation of boiling. From the study, it is concluded that if the sodium void reactivity is limited (4.6$) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters and also it is found out higher primary pump flow halving time (15 s instead of 8 s) can avoid cliff edge effects in 1000 MWe MFBR transient behaviour

  6. Containment design requirements and their application to 500 MWe plant

    Containment is the final barrier to the release of fission products from the reactor system to the environment. In the defence in depth philosophy to reactor safety, the containment is one of the four special safety systems. Therefore, comprehensive requirements are specified by Atomic Energy Regulatory Board (AERB) in its design code of practice. The code stipulates that in case of double containment, the secondary containment should completely envelope the primary. Further, the annulus space between the primary and secondary containment envelope shall be provided with a purging arrangement to maintain a negative pressure in the space thus ensuring zero ground level release consistent with ALARA principles. In this presentation the various AERB requirements and how these requirements have been met in 500 MWe containment design are discussed. Also containment response to some major accident types have been briefly described. (author). 4 refs., 2 figs

  7. Primary circulating pump trip transient analysis for 500 MWe PHWR

    The 500 MWe Indian pressurised heavy water (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like primary heat transport (PHT) system configuration with two loops, four primary circulating pumps (PCPs) and four passes through core, addition of a pressuriser (surge tank) in the PHT system along with feed/bleed system and their safety related implications, simulation model development and transient analysis studies are necessary. The paper deals with the details of the mathematical model for PHT system and parametric study on one PCP trip transient analysis with set/step back. The studies were carried out after including the proposed new SGPC program similar to 220 MWe PHWR, which gives a 48 kg/cm2 maximum SG pressure setpoint at zero power, without changing the 100% power set pressure

  8. Advanced Structural Mechanics Design of 500MWe Commercial SFRs

    In future 500 MWe SFRs, innovative design features have been incorporated in the reactor assembly components, to achieve improved economy and enhanced safety. The major design changes are: (1) innovative configuration of main vessel bottom dished head, (2) dome shaped roof slab with conical support, (3) thick plate for the rotatable plugs instead of box type structure, (4) welded grid plate with reduced number of sleeves, reduced diameter of intermediate shell and reduced height, (5) increased number of seamless primary pipes, (6) inner vessel with single radius torus welded with the grid plate, (7) integrated liner and safety vessel with thermal insulation arrangement, (8) innovative core support structure, (9) introducing in-vessel purification (10) integration of control plug and small rotatable plug and (11) straight pull fuel handling system. These features also help to shorten the construction time significantly. To validate the innovative design features, structural analysis for the geometrical optimization, investigation of buckling of thin shells, integrity assessment of integrated reactor assembly components were carried out. Further, to demonstrate the manufacturing feasibility of the new designs, technology development activities have been completed successfully. These developments are (i) thick tri junction forging for dome shape roof slab, (ii) welded grid plate, (iii) thick plate narrow gap welding for rotatable plugs, (iv) doubly curvature inner vessel sector, (v) embedded safety vessel with thermal insulation panel and (vi) large diameter bearing. The presentation brings out clearly the main structural features of the innovative concepts, which have been incorporated in future designs, application of advanced structural mechanics analyses carried out to comply the RCC-MR (2010), design requirements, challenges and achievements of technology development exercises completed. (author)

  9. Dynamic modelling of pressure control system of a 500 MWe PHWR power plant thermal hydraulics aspects

    A computer code for the dynamic analysis of the proposed 500 MWe Pressurised Heavy Water Reactor is being developed. One of the modules of this code deals with the primary heat transport system pressure control. A thermal hydraulic model of the pressure control system has been developed. This model includes the following : reactor coolant loop, primary circulating pump, Core heat transfer, Feed/bleed with Bleed Condenser and pressure controller. Analysis has been carried out for transients like change in reactor power, leakage from the primary heat transport system and malfunctioning of control system. The mathematical model is presented in the paper along with the results obtained for some of the transients analysed. (author). 6 refs., 6 figs

  10. Ergonomic design of mosaic control panel and standardised control tile configurations for 500 MWe PHWR

    A review of control rooms of operating nuclear power plants identified many design problems having potential for degrading the performance of operators. Many indications and controls on existing control panels are placed outside the recommended visual and reach envelopes for acceptable operator usage. As a result, the application of human factor principles was found to be needed. This paper describes the design approach for working out the dimensions of main control room panels and console using human engineering principles and recommends the ergonomic dimensions of the main control room panels and console. Further it gives the basis and works out the control tile configurations for 500 MWe PHWR project. It also suggests the use of a full scale mock up for design evaluation and verification. (author). 7 refs., 4 figs

  11. Stability analysis of through wall cracked primary heat transport pipe of 500 MWe PHWR - Part 1

    The advent of Leak-Before-Break (LBB) concept is progressively replacing the traditional design basis event of Double Ended Guillotine Break (DEGB) in the design of high energy fluid piping system. The stability analysis of the through-wall cracked primary heat transport pipe of 500 MWe PHWR is carried out by J-integral (J) and tearing modulus (T) concept. The flaws are assumed in circumferential and longitudinal directions. The loadings considered are bending moment due to Safe Shutdown Earthquake (SSE) and axial force due to the pressure of the coolant. The critical size of the circumferential flaw which leads to catastrophic failure is determined under the assumed loading conditions. The leak rate is determined based on LEFM with Irwin's plastic correction. The leakage size crack is determined by applying margin of 10 on detectable leak rate. The crack stability is checked for leakage size crack under normal plus seismic stresses. (author). 29 refs., 19 tabs

  12. Fire protection in nuclear power plants of 500 MWe (Paper No. 1.7)

    Fire protection system for the 500 MWe TAPP - 3 and 4 units caters to the requirement of fire prevention in various plant buildings. The systems consists of fire water system, gaseous fire extinguishers system, fire detection and alarm system and portable fire extinguishers. The fire water system is designed as per the guidelines of USNRC1-120, IAEA Safety Guide Series - 50-SG-D2, NFPA Codes, TAC as well as IS:1641 to 1648. Dedicated storage of water as per the code is provided in a seismically designed reservoir on which fire pumps are also installed. Adequate number of pumps are provided taking into consideration the codal design criteria. In the unlikely event of the requirement of fire water extending beyond design durations, water from another reservoir having a large capacity can be transferred to the reservoir encompassing dedicated fire water storage. During the remote occurrence of a station black-out, (when neither off-site sources nor on-site emergency sources of power supply is available), provision have been incorporated to supply water to the Steam Generators and the End Sheilds to remove the decay heat, through a diesel engine driven fire water pump. At RAPP, a unitised fire water system is proposed to be designed for each unit as seismically designed Induced Draft Cooling Tower for each unit have been provided which cater to the requirement of fire water storage also. In order to prevent the damage due to spray of water in certain areas and areas which are enclosed like computer room in control building. Electronic Data Processing Centres and false floor below Computer Room and main Control Room, Gaseous Fire Extingushing System using halogenated hydrocarbons (Halon-1301) is provided. The approaches considered for the design of the fire protection system for 500 MWe nuclear plants with particular reference to TAPP-3 and 4 units have been discussed in this paper. (author)

  13. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs

  14. Final report on the evolution of supporting conditions for the feeders of 500 MWe PHWR

    This report deals with the evolution of generic supporting conditions for the feeders of 500 MWe PHWR based on the analysis and qualification of a few representative feeders. There are 196 different feeder pipe configurations for a total of 748 feeders. The present analysis was aimed at evolving a generalised supporting criteria based on the analysis of some representative feeders. The analysis was carried out for various loadings viz. pressure, temperature, dead weight, operating basis earthquake (OBE), safe shutdown earthquake (SSE) and creep loadings. The analysis for OBE and SSE loadings were carried out using response spectrum method. The effect of spacers between various feeders was modelled using higher damping values than those prescribed in ASME code. Based on the above analyses, generic supporting arrangements for the feeders of various groups have been finalized. This report gives details about the mathematical modelling, the analysis approach, the optimised supporting criteria, finalization of grouping and fixing of boundaries between various groups of feeders. (author). 34 refs., 51 figs., 69 tabs

  15. Evolution of supporting conditions for some representative feeders of 500 MWe PHWR

    This report deals with the analysis and qualification of some representative feeders of 500 MWe PHWR. There are 196 different feeder pipe configurations for a total of 748 feeders. Analysing these many configurations is a difficult task. Therefore, the present analysis was aimed at evolving a generalised supporting criteria based on the analysis of some representative feeders. For this purpose, these feeder pipes were classified into 4 groups based on their geometrical configurations. A total of 7 feeders were selected, which broadly represent the feeders of all the 4 groups. Many of these feeders are interconnected by spacers, which make their analysis for earthquake loadings quite complex. The analysis was carried out for various loadings viz., temperature, pressure, dead weight and operating basis earthquake (OBE). The analysis for OBE loading was carried out using response spectrum Method. Based on the analysis of 7 feeders, generic supporting arrangements for feeders of various group have been suggested. The report gives in detail the input data used, mathematical modelling, analysis approach, various supporting criteria evolved, the optimised supporting criteria and the conclusions arrived at. (author). 20 refs., 18 figs., 28 tabs., 1 appendix

  16. Structural safety analysis of 500 MWe pressurized heavy water reactor containment for air craft impact load

    Full text: Transient dynamic analysis of 500 MWe PHWR nuclear containment has been carried out to evaluate the damage that may be caused due to aircraft impact of Boeing 707-320. A special three dimensional finite element analysis procedure has been developed for the double containment walls with an annulus gap in between outer containment wall (OCW) and inner containment wall (ICW). The case studies include the analyses of OCW single model and the combined model of OCW and ICW for impulsive load due to the aircraft impact. It is demonstrated that OCW would suffer local perforation with a peak local deformation of ∼94 mm without loss of the overall integrity. However, this first barrier cannot absorb the full impulsive load. Here after the local perforation of the first barrier OCW, the impulse load is transferred to the second barrier ICW in the combined model of OCW and ICW. In the analyses of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.195 sec due to the local perforation of OCW. This results in the local deformation of the ICW ∼66 mm. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls for aircraft impact with higher energies

  17. Stress analysis of 500 MWe PHWR calandria non-radial nozzles for external loadings

    The 500 MWe Pressurised Heavy Water Reactor (PHWR) calandria is a stainless steel horizontal cylindrical vessel which houses the core of the reactor. It has a number of vertical non-radial nozzles which locate shut down, reactivity control and over pressure relief safety devices. The calandria is designed, fabricated and inspected as per ASME Boiler and Pressure Vessel Code class one component stringent requirements. Model studies of this equipment is carried out to know the stress concentration factors for various thermal and seismic loads transferred through these nozzles. In this experiment nozzles lying in one quadrant of calandria are simulated on a 1:3.2 size model and six types of forces are applied on each nozzle. Total 136 numbers of 3 element strain rosettes have been used in this experiment. The strain data are collected through a 100 channel data recording system and stresses on prototype calandria are predicted for various load cases with the help of a BASIC language rosette analysis program implemented on the microprocessor system of the data logger. The interaction effect between various nozzles is also studied. Comparisons have been made between present experimental results and simplified analytical design calculation (based on Bijlaard's method) along with finite element analysis results of a selected group of 3 nozzles. It is shown that local stresses near an opening decay within a short distance from root of the nozzle and interaction effect on neighbouring nozzles surrounding a loaded nozzles is small. The stress concentration factor data generated can be used by the designer to calculate peak and local stresses on the calandria shell-nozzle junctions of various reactivity devices for all six kinds of nominal thermal and seismic loads. (author). 3 refs., tabs., figs

  18. The transuranic mass balance during the introduction of metal fuel FBR cycle

    The mass flow of plutonium and minor actinides is calculated for a future light water reactor-fast breeder reactor (LWR-FBR) transition scenario, in which a certain scale of power generation by LWRs is continued for a long period before the replacement by FBRs begins. The burnup of the LWR spent fuel is considered to be higher than the current standard. It is assumed that all the plutonium and minor actinides recovered from LWRs are used for the start-up and feed of metal fuel commercial FBRs to replace those LWRs that reach the lifetime. The results show that the accumulated plutonium and minor actinides from the LWRs can be consistently consumed without further accumulation, by gradually establishing the same scale of the FBR power generation and the associated fuel cycle. The minor actinides content in the standard FBR fuel that is sufficient to realize this scenario is estimated to be about 2wt%. This result indicates that if the FBR era is to come in the future, the extended LWR era causes no significant problem in terms of the consumption of the accumulated transuranic. (author)

  19. Condenser circulating water system for 500 MWe units (Paper No. 7.7)

    Condenser Circulating Water (CCW) system for Tarapur Atomic Power Project (TAPP), is a once-through system using sea water for cooling, envisaging a flow of 115,000 M3/hr to dissipare a condenser heat load of 1.062x10 kcal/hr per unit to the sea. It is proposed to use pumps of 30,000 M3/hr developing 17.2 MWC head. Such large capacity pumps are to be manufactured indigenously for the first time. Condensers with titanium tubes and titanium clad MS tube sheets are proposed to be used keeping in view the corrosive and erosive nature of the sea water. On load tube cleaning system (OLTCS) using sponge balls and debris filters located on the inlet of the condensers are to be used to maintain low fouling factors in the condenser tubes. Based on techno-economic comparison between the steel conduits and RCC tunnels, it has been proposed to use RCC tunnels for the CCW system. Hot water outlet from the condenser and other auxiliary systems is led through a specifically designed discharge structure. Model studies were carried out at Central Water and Power Research Station (CWPRS), Pune for the various discharge configuration in order to achieve the stipulated temperature differentials. In order to handle the large amount of cooling water, centre flow rotating band type screens are considered on the inlet of pumps. These types of screens though very popular in power station abroad, are yet to be installed in any power station in India. The philosophy of unitised concepts has been adopted and each unit is fed from an independent CCW system. To safeguard against the corrosive nature of sea water all pumps internals are proposed to be manufactured with SS 316 and all portions of the steel pipe lines shall be internally lined with cement mortar. The design of the CCW system has been optimized based on minimum total evaluated cost. The paper highlights the salient features of the CCW system designed for the 500 MWe Nuclear Plants being set up at TAPP and RAPP. (author)

  20. A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

    Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.

  1. A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

    Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis

  2. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  3. Reactor Physics and Safety Aspects of Metal Fuelled FBR

    It is known that under the Beyond Design Basis Accident (BDBA) of Unprotected Loss of Flow (ULOF), the oxide fueled reactor shows early onset of sodium boiling and fuel melting leading possibily to near prompt criticality and entry into the disassembly phase in less than 100 s. Higher total Na void reactivity effect and lower Doppler effect may be expected to make the accident scenario severe in metal core also. However, from studies reported here, ULOFA is found to be benign for the metal fueled reactor as the sodium coolant boiling is delayed beyond 900 s. This behaviour is due to lower fuel temperatures (close to that of coolant) and higher negative expansion feedbacks. The reactor becomes subcritical and the principal heat source after 600 s is the fuel radioactivity decay heat. An accurate modeling of decay power is important when dealing with the transient at long times. If fuel decay heat removal systems are designed to reliably remove decay heat, after ULOFA, the metal fuelled reactor reaches a safe subcritical state at a elevated Na temperature. (author)

  4. Seismic analysis of fuelling machine bridge, carriage and column assembly of 500 MWe PHWR (in unclamped conditions)

    This report deals with the seismic qualification of fuelling machine (FM) bridge, carriage and column assembly of 500 MWe Pressurised Heavy Water Reactors (PHWRs). FM bridge carriage and column assembly along with the fuelling machine head (FMH) performs the important task of on-power refuelling which is an essential feature of all PHWRs. Two fuelling machine heads home on to the opposite ends of a coolant channel to perform the refuelling operation. It is essential to refuel 2 channels for a full power day of a reactor. The fuelling machine can clamp onto 392 possible channel locations for a 500 MWe reactor. Hence it becomes very important to seismically qualify the fuelling machine since it becomes part of the pressure boundary during refuelling process. Because of the close proximity of FMH to the channels, it becomes necessary to seismically qualify the fuelling machine even in the unclamped condition. Present report gives the analysis for the uncoupled condition of FMH under both operating base earth quake (OBE) and safe shut-down earthquake (SSE) conditions using the response spectrum method. Based on the analysis it has been found that the components of FM bridge, carriage and column assembly can withstand the seismic forces generated and the stresses are within their allowable limits both under OBE and SSE loadings. This report gives in detail the input data used, mathematical modelling various boundary conditions, results and the conclusions arrived at. (author). 12 refs., 7 tabs., 14 figs., ills

  5. Modelling and analysis of the transient response of the steam generator of a 500 MWe PHWR power plant

    As a part of the development of a power plant dynamic analysis code for the 500 MWe pressurised heavy water reactor (PHWR), a computer code for the analysis of the behaviour of U-tube natural circulation steam generator (SG) has been developed. All the associated control systems are also modelled. This module is integrated with the dynamic analysis code for 500 MWe PHWR and used for the simulation of transients. The model is based on the one dimensional, non-linear, single fluid conservation equations for mass, momentum and energy. An empirical slip flow model is included to enable description of non-equilibrium two-phase flow. The model accounts for both fluid compressibility and thermal expansion effects. This is achieved by including the implicit energy dependence in the coupled equations of mass, momentum and fluid state, and by solving the full system of fluid equations through a two-step iterative technique. Optimisation studies related to PHT pressure control systems are being carried out with the integrated code. The objective of present studies is to investigate the variation in the PHT system behaviour following severe anticipated operational transients such as turbine trip. The details of the mathematical model of the steam generator as well as system response following turbine trip are discussed in the paper. (author). 8 refs., 8 figs

  6. The questions of liquid metal two-phase flow modelling in the FBR core channels

    The two-fluid model representation for calculations of two-phase flow characteristics in the FBR fuel pin bundles with liquid metal cooling is presented and analysed. Two conservation equations systems of the mass, momentum and energy have been written for each phase. Components accounted the mass-, momentum- and heat transfer throughout the interface occur in the macro-field equations after the averaging procedure realisation. The pattern map and correlations for two-fluid model in vertical liquid metal flows are presented. The description of processes interphase mass- and heat exchange and interphase friction is determined by the two-phase flow regime. The opportunity of the liquid metal two-phase flow regime definition is analysed. (author)

  7. Structural dynamics in FBR

    In view of thin walled large diameter shell structures with associated fluid effects, structural dynamics problems are very critical in a fast breeder reactor. Structural characteristics and consequent structural dynamics problems in typical pool type Fast Breeder Reactor are highlighted. A few important structural dynamics problems are pump induced as well as flow induced vibrations, seismic excitations, pressure transients in the intermediate heat exchangers and pipings due to a large sodium water reaction in the steam generator, and core disruptive accident loadings. The vibration problems which call for identification of excitation forces, formulation of special governing equations and detailed analysis with fluid structure interaction and sloshing effects, particularly for the components such as PSP, inner vessel, CP, CSRDM and TB are elaborated. Seismic design issues are presented in a comprehensive way. Other transient loadings which are specific to FBR, resulting from sodium-water reaction and core disruptive accident are highlighted. A few important results of theoretical as well as experimental works carried out for 500 MWe Prototype Fast Breeder Reactor (PFBR), in the domain of structural dynamics are presented. (author)

  8. Seismic analysis of two heavy water upgrading towers for 500 MWe Tarapur Atomic Power Plant-3 and 4

    The report deals with the analysis carried out for the evaluation of earthquake induced stresses and deflections in two 1500 mm diameter heavy water upgrading towers for Tarapur Atomic Power Plant-3 and -4. The analysis of upgrading towers has been carried out for two mutually perpendicular horizontal excitations and one vertical excitation applied simultaneously. The upgrading towers have been analysed using beam model taking into account soil-structure interaction. Response spectrum analysis has been carried out using envelope spectra for 500 MWe sites. The seismic analysis has been carried out for the towers with supporting structure along with concrete pedestals and raft foundation. The towers have been checked for their stability due to compressive stresses to avoid buckling so that safety of the nearby structures is not damaged even in the event of SSE (Safe Shutdown Earthquake) loading. (author). 16 refs., 11 figs., 18 tabs

  9. Transient thermal hydraulic behaviour of the fuel bundles during on-power unloading operation in the proposed 500 MWe PHWR

    One of the main objectives under design and development of fuel in water cooled nuclear reactors is to ensure fuel integrity during spent fuel handling operation. The on-power refuelling facility adopted in the Indian Pressurized Heavy Water Reactors (PHWRs) causes exposure of the irradiated fuel, during its unloading, to wide variations in its surroundings including exposure to dry gaseous environment. Detailed analyses have been carried out to assess the fuel pin temperature transients during the entire course of its passage from within the reactor to the outside surroundings to ascertain fuel integrity. The cases of normal as well as envisaged off-normal transport operations have been considered in these calculations. The forced air cooling provisions have also been worked out to mitigate the consequences of off-normal transport operation. The present paper deals briefly with the system description, method of calculations and the results obtained for the case of spent fuel handling in the proposed 500 MWe PHWR. (author)

  10. Elasto-plastic finite element analysis of axial surface crack in PHT piping of 500 MWe PHWR

    The leak before break (LBB) approach in nuclear piping design envisages demonstrating that the pressurized pipe with a postulated flaw will leak at a detectable rate leading to corrective action well before catastrophic rupture would occur. This requires analysis of cracked pipe to study the crack growth and its stability. This report presents the behaviour of a surface crack in the wall of a thick primary heat transport (PHT) pipe of 500 MWe Indian PHWR. The line spring model (LSM) finite element is used to model the flawed pipe geometry. The variation of crack driving force (J-integral) across the crack front has been presented. The influence of crack geometry factors such as depth, shape, aspect ratio, and loading on peak values of J-integral as well as crack mouth opening displacement has been studied. Several crack shapes have been used to study the shape influence. The results are presented in dimensionless form so as to widen their applicability. The accuracy of the results is validated by comparison with results available in open literature. (author). 47 refs., 8 figs

  11. Development of high pressure conductivity probe (HPCP) for secondary shut down system (SDS-2) of 500 MWe PHWR

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface. This interface moves towards the calandria because of molecular diffusion, temperature difference and physical disturbances in the moderator level. It is proposed to install two numbers of high pressure conductivity probes (HPCP) to monitor the interface movement as well as to provide the safe annunciation value for interface location. On actuation of the SDS-2 signal, high-pressure helium will inject the poison into the moderator to shutdown the reactor. During poison injection, these probes will experience high pressure of nearly 85 kg/sq.cm. Global market survey indicated that conductivity probes having built in temperature sensor are available for a maximum pressure rating of 35 kg/sq.cm. Hence in order to meet the process requirement of SDS-2, the development of HPCP suitable for a pressure of 85 kg/sq.cm. was taken up. Two numbers of such probes were successfully designed, fabricated and evaluated for their performance. The developed conductivity probes fully meet the laid design and performance criteria. The aforesaid development work was a successful endeavour towards indigenisation of high-pressure conductivity probe for future applications. This report deals with the design aspects, fabrication technique, material and performance evajuation criteria and test results of HPCP. (author)

  12. Stress analysis of secondary ramp and secondary tilting mechanism of inclined fuel transfer machine for 500 MWe PFBR

    Inclined Fuel Transfer Machine (IFTM) is one of the important machine of the fuel handling system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It is used to transfer core sub-assemblies (CSA) from reactor vessel to fuel building and vice-versa. Secondary ramp and Secondary tilting mechanism (SR/STM) is a part of IFTM which acts as a passage to transfer CSA. This mechanism and components were designed by the Refuelling Technology Division of BARC as per the ASME design code as class 2 component. Being critical in nature and complicated in geometry it was required to check the design of these components by detailed finite element analysis. The loading considered in the present study was static, thermal and seismic conditions. This was done using FEM software COSMOS/M. The Stresses were categorised as per the requirement of the ASME code for various levels of loading (Level A, B and C). Based on the analysis performed, it was concluded that the SR/STM qualifies the requirement of ASME code Section-III NC (Class-2 components). This report gives the details of the studies performed. (author)

  13. Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

    Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

  14. The basic research on the CDA initiation phase for a metallic fuel FBR

    Hirano, Go; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan); Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  15. Design and manufacture of tube to tubesheet joints of steam generator for 500 MWe Prototype Fast Breeder Reactor

    Prototype Fast Breeder Reactor (PFBR) is 500 MWe pool type sodium cooled fast reactor. Presently this reactor is at advanced stage of construction at Kalpakkam. The main function of the steam generator is to extract the reactor heat through secondary sodium system and convert the feed water into superheated steam in the tubes of steam generators. The steam generator is a vertical shell and tube type heat exchanger with liquid sodium in the shell side and water/steam in the tube side. Operating experience of FBRs have shown that steam generator (SG) holds the key to commercial success of such reactors. Tube leakage is a serious problem and the prevention of sodium water reaction incident in the SG is essential to maintain the plant availability. In case of crack/failure in tube, high pressure water/steam reacts with shell side sodium and results in exothermic reaction with evolution of hydrogen, corrosive reaction products and intense local heat depending on leak size. This high reactive nature of sodium with water/steam requires that sodium to water/steam boundaries of steam generators must possess a high degree of reliability against failure. This is achieved in design and manufacturing by maximising the tube integrity and more importantly by proper selection of tube to tubesheet joint configuration. The principal material of construction of SG is Modified 9Cr-1Mo steel. The tubes are seamless and produced by electric arc melting followed by Electro Slag Refining (ESR) with tight control on inclusion content. Ultrasonic and eddy current testing is done on entire tube length in accordance with ASME SEC III Class I. Long seamless tubes (each 23m) are used in order to reduce the number of tube to tubesheet welds.Each SG has 547 tubes and there are 9 SG in the reactor including one spare module. There is no tube to tube joint as the aim is to minimise the number of welds to increase reliability.Tube to tubesheet joint selected for PFBR steam generator is of internal

  16. Leak-before-break qualification of 500 MWe PHWR PHT straight pipes by J-integral-tearing modulus and limit load method

    The concept of leak-before-break (LBB) has now-a-days replaced the traditional design basis event of double ended guillotine break (DEGB) to design the primary heat transport (PHT) piping system of the future generation nuclear reactors. Consequently, the LBB approach is adopted in the design of PHT system of India's future generation 500 MWe nuclear reactors. The LBB approach aims at the application of fracture mechanics principle to demonstrate that pipes are, in general, unlikely to experience DEGB without prior indication of leakage. It shows that a through wall leakage size cracks (LSC) in the pipe is stable under the maximum credible loading condition. This is to be shown for all the piping components, namely, straight pipes, elbows and branch tees in the entire PHT system. The present report details the LBB qualification of the straight pipe portions of the 500 MWe PHT pipe layout. The qualification is done through stability analysis of the pipes with postulated LSC by J-integral-tearing modulus and limit load method. The report has been split into five sections and two appendices. Section 1 describes the general methodology of the LBB analysis. Section 2 describes the evaluation of J-integrals by analytical estimation schemes. Section 3 details the finite element analysis of the pipes with postulated cracks to evaluate J-integral, limit load and crack opening area. Section 4 shows the critical loads evaluated by J-T method. Section 5 demonstrates the LBB qualification of the pipes by showing the necessary factors of safety. Appendix A describes in brief the analytical J-estimation schemes and appendix B compares the two methods to calculate tearing modulus. (author)

  17. Integration of functional reliability analysis with hardware reliability: An application to safety grade decay heat removal system of Indian 500 MWe PFBR

    A passive system can fail either due to classical mechanical failure of components, referred to as hardware failure, or due to the failure of physical phenomena to fulfill the intended function, referred to as functional failure. In this paper a methodology is discussed for the integration of these two kinds of unreliability and applied to evaluate the integrated failure probability of the passive decay heat removal system of Indian 500 MWe prototype fast breeder reactor (PFBR). The probability of occurrence of various system hardware configurations is evaluated using the fault tree method and functional failure probabilities on the corresponding configurations are determined based on the overall approach reported in the reliability methods for passive system (RMPS) project. The variation of functional reliability with time, which is coupled to the probability of occurrence of various hardware system configurations is studied and incorporated in the integrated reliability analysis. It is observed that this consideration of the dependence of functional reliability on time will give significant advantages on system reliability. The integrated reliability analysis is also explained using an event tree. The impact of the provision for forced circulation in the primary circuit on functional reliability is also studied with this procedure and it is found that the forced circulation capability helps to bring down the total decay heat removal failure probability by lowering the peak temperatures after the reactor shut down.

  18. Physics Design and Safety Studies of a 320 MWt Experimental Metal FBR

    Fast reactors play a major role in the sustained growth of nuclear power in India. Metal fuelled fast reactors with high breeding ratio are essential to meet the rapid growth rate of energy demand. Several R&D programs have been initiated at IGCAR for establishing this technology with closed metal fuel cycle. Construction of a Metal Fuel Test Reactor (MFTR) at IGCAR has been planned to enable full-scale experimental testing of fuel sub-assemblies of a typical commercial power reactors before their launching. With the availability of MFTR, it is possible to experimentally validate various core physics design parameters of metal fuels. There would be a co-located pyro-reprocessing plant to demonstrate closing the fuel cycle. The discharged fuel from MFTR will be recycled after reprocessing and fuel fabrication. It is, therefore, desirable to have high burnup for mastering reprocessing of high burnup metal fuel sub-assemblies. MFTR will also be used to produce radio isotopes for medical applications and also for the development. Irradiation experiments towards validation of new fuel and structural materials, capable of achieving higher burnup and doses, are also planned in MFTR. In order to meet these objectives, the physics design of a MFTR core with breeding ratio ~1.1 is performed using sodium bonded ternary fuel (U-Pu-6%Zr). Core height is fixed to be 1000 mm. Design is performed to achieve a peak linear heat rating (LHR) of 450 W/cm and peak burnup of 150 GWd/t. Minimum power level satisfying the above requirements is found to be 320 MWt. Salient features of this core is discussed in this paper. (author)

  19. Development of innovative reactor assembly components towards commercialization of future FBR

    PFBR, which is under construction in India, is a 500 MWe power, sodium cooled, (U-Pu)O2 fuelled, pool type fast reactor. Beyond PFBR, it is planned to construct 3 twin units; each one is 2 x 500 MWe capacity sodium cooled fast reactors with improved economy and enhanced safety. Significant capital cost reduction is targeted by way of introducing innovative and improved concepts for the reactor assembly components, such as grid plate, primary pipes, top shield and fuel handling system and optimizing the main vessel diameter and bottom dished head shape. The improvements have been conceptualized based on preliminary analysis and the detailed analysis and development for few of the component have been undertaken. The capital cost reduction of the reactor assembly components that could be achieved through these improved concepts is estimated to be about 25 %. These apart, the safety of the reactor is enhanced by passive features in shutdown and decay heat removal systems. The innovations introduced have many unique features and it would form an international bench mark for the future FBRs. To validate these concepts, R and D areas have been identified and strategy to execute the R and D has been defined clearly. Towards commercialization of future FBR, special efforts are put forth to manufacture the components of the entire reactor assembly, which are non-replaceable, as a factory-made single item. In this paper, the basis of each concept to depict the Indian approach and strategy to make the fast reactor economically competitive is highlighted. (author)

  20. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase

  1. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    Riyas, A., E-mail: rias@igcar.gov.in [111B, CDO, Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Mohanakrishnan, P. [Adjunct Professor, Manipal University, Manipal (India)

    2014-10-15

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase.

  2. FBR type reactor

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  3. coustic Leak Detection Based on Wavelet Packet and Genetic Algorithm for LM FBR Steam Generators

    2011-01-01

    Steam generator is one kind of key equipments in liquid metal fast breeder reactors (LM FBR) whose reliability will influence the safety of nuclear power plant. We can see that SG is the highest risky equipment from the running experience

  4. Evaluation of the commercial FBR introduction date

    This report examines one criterion for introducing a commercial FBR: economic competitiveness with a Light Water Reactor (LWR). For this analysis, the commercial FBR is assumed to be the fifth-of-a kind replicate which represents an economically mature plant. This FBR is deemed economically competitive when its life-cycle energy cost is less than or equal to that of an LWR. Results of this analysis are used in a comparative analysis of alternative FBR development stategies. The strategies evaluated in these studies assume both 1000- and 1457-MWe FBRs. Since the capital costs per kilowatt, and therefore the energy costs, for these two FBR sizes are different, they will become economically competitive at different times. The probability density function for the 1457-MW(e) FBR has an expected value date or weighted average date of 2030, compared with 2033 for the probability density function for the 1000-MW(e) FBR

  5. Design study and evaluation of advanced fuel fabrication systems for FBR fuel cycle

    The conceptual design study for advanced FBR fuel fabrication system has been performed for the purpose that the feature of small-scale fabrication system in the transition stage from LWR to FBR fuel cycle. On the small-scale of 50 ton heavy metal per year fabrication system, dry type fabrication systems have superior cost performance than the wet type, although waste amount is larger. (authors)

  6. Inspection of FBR using virtual reality

    The under sodium viewer (USV) employing ultrasonic techniques can obtain images of structures in a fast breeder reactor (FBR) though the FBR is filled with opaque liquid sodium as coolant. Even using the USV, however, operators have been unable to directly look around the internal structures of an FBR because of the long time required to form images from the reflected ultrasonic waves. As a result, it has been difficult to appropriately control the positioning of the USV. We have constructed a virtual FBR in which the sodium becomes transparent, to enable an operator to inspect the internal structures. This paper reports on the conditions and supporting method that facilitate three-dimensional operation of the USV in the virtual FBR. (author)

  7. Summary: analysis of alternative FBR development strategies

    This report summarizes the comparative evaluation of alternative strategies for the development of the commercial fast breeder reactor (FBR) in the United States. For planning purposes, a range of possible FBR development paths called strategies were selected for evaluation. These strategies, designed to be technically and economically feasible, were expressed in terms of the timing and nature of facilities/research and development programs required to reach full power operation of the first commercial FBR. Four of the seven strategies resulted in a large (1457 MWe) FBR as an end point, the other three in a 1000-MWe plant. Probability distributions were calculated for total strategy costs and time to completion. For the seven strategies analyzed, the costs (discounted 1980 dollars) ranged from $1.8 billion to $4.9 billion; the completion times ranged from 24 to 55 years

  8. Scenario study on the FBR deployment

    This study on success scenarios for the Fast Breeder Reactor (FBR) deployment was performed taking account of future situation of fossil, renewable and nuclear energies in Japan as well as the world from the viewpoints of the following four items; economics, environment, energy security and restriction of natural uranium resources. In the economics scenario, if carbon tax is added to generating cost of LNG, coal and oil and the economics of FBR cycle is competitive with LWR cycle in the future, FBR cycle will be expected to introduce as the middle and base load power plant. In the environment scenario, there is also any possibility that FBR cycle which can burn and transmute minor actinide and fission product elements will be introduced in order to reduce the burden of deposit facility and the toxicity of high-level waste. In the uranium resources restriction scenario, FBR cycle needs to be deployed at the latest in the middle of 21st century from the viewpoint of the restriction of natural uranium resources. This study was carried out in a part of JNC's feasibility study on commercialized FBR cycle system. (author)

  9. Creep fatigue design of FBR components

    This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

  10. Seismic stability assessment for FBR main vessels

    Buckling under earthquake loading is one of the most important problems for thin-walled structures such as FBR main vessels. Therefore, Central Research Institute of Electric Power Industry (CRIEPI, Japan), commissioned by the Ministry of International Trade and Industry of the Japanese Government, has carried out the demonstration test and research program of buckling of FBR. The objectives of this paper is to describe the results of static and dynamic buckling tests and numerical analyses of cylindrical shells under horizontal loads, and to present a seismic stability assessment method for FBR main vessels. Shear-bending buckling strength evaluation formulae and response reduction factors due to nonlinearity in pre-buckling stage are worth special mention

  11. Characterization of alternative FBR development strategies

    Near-term decisions regarding the nature and place of the FBR development program must be made. This study is part of a larger program designed to provide the Department of Energy (DOE) with imformation that can be used to make strategic programmatic decisions. The focus of this report is the description of alternative approaches for developing the FBR and the quantification of the duration and cost of each alternative. The time frames of the alternative approaches are investigated in companion reports (White 1981 and Fraley 1981). The results of these analyses will be described in a summary report

  12. Advanced Structural Mechanics Design of 500 MWe Commercial SFRs

    • PFBR design, manufacture, construction and safety review have given rich experience. • Comprehensive roadmap has been drawn to design and develop future FBRs with focus on economy and standardisation. • Advance detailed Structural analysis for various structure were carried out to understand the behaviour of the components

  13. FBR and RBR particle bed space reactors

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 100K), high coolant-outlet temperatures (1500 to 30000K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H2-cooled mode. The RBR will operate only in the open-cycle H2-cooled mode

  14. Part 6. Internationalization and collocation of FBR fuel cycle facilities

    This report examines some of the non-proliferation, technical, and institutional aspects of internationalization and/or collocation of major facilities of the Fast Breeder Reactor (FBR) fuel cycle. The national incentives and disincentives for establishment of FBR Fuel Cycle Centers are enumerated. The technical, legal, and administrative considerations in determining the feasibility of FBR Fuel Cycle Centers are addressed by making comparisons with Light Water Reactor (LWR) centers which have been studied in detail by the IAEA and UNSRC

  15. Device for reducing radioactive corrosion product in FBR type reactor

    The present invention concerns an FBR type reactor using liquid metal as coolants, connecting the reactor core with a heat exchanger by way of cooling system pipeways and recycling the coolant by the driving force of a pump. A bypass circuit is disposed to a portion of a cooling system, and a vessel inserted with fillers is disposed to a portion of the bypass circuit. The coolants are prepared with the same material as that for the reactor core constituent material. The filler suffered from corrosion with sodium coolants and to increase the concentration of the corrosion products in sodium. This suppresses the corrosion of nuclear fuel cans in the reactor core. Accordingly, leaching of radioactive corrosion products such as Mn or Co caused by the reduction in the wall thickness of the fuel can can be suppressed. (I.J.)

  16. Development of FBR cycle data base system

    In the 'Feasibility Study on Commercialized Fast Reactor Cycle System (F/S)'. scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show significance of the Fast Breeder Reactor (FBR) cycle system introduction concretely are performed in parallel with a design study for FBR plants, reprocessing systems and fabrication systems. In these evaluations, informations such as economic prospects, prospects for supply and demand of resources and a progress of engineering development are used in addition to design information. This report explains a FBR Cycle Database in order to carry out management and search of various design information and the relating information. The prototype system of the database was completed in the 2000 fiscal year, and the problem of the user number restriction of the prototype system has been improved by Web-ization in the 2001 fiscal year. About 7,000 data are stored in this data base (as of the end of March, 2002). The expansion of user etc., and the continuation of input work of various evaluation information will be carried out, in the phase 2 of F/S. (author)

  17. Experimental and computational study on prediction of natural circulation in top-entry loop-type FBR

    The present paper deals with a methodology to predict decay heat removal capability through natural circulation in a liquid metal-cooled fast breeder reactor (FBR) in both experimental and computational methods. Systematic discussion is given to derive suitable similitude conditions for an experiment of a scaled water model. The validity of the derived similitude conditions is estimated by comparing experimentally predicted values with computed values of typical physical quantities such as temperature and flow rate. The results show that the experimental prediction is of practical use in FBR designs as well as the computational one. (author)

  18. Important matters in realizing commercial FBR cycle

    Full text: Nuclear energy, one of the essential measures against increasing world energy demand and global warming is expected to expand in near future. The Fast Breeder Reactor (FBR) cycle technology is supposed to enhance the utilization of uranium resources for the long term. Many countries are interested in pursuing the technology development. Japan with scarce natural resources also has been developing the technology for decades, since its beginning of utilization of nuclear energy. As a person who spent many years in the fields of nuclear power generation, I sincerely hope that you, young generation will work hard passionately and succeed in commercializing the FBR cycle. I would like to show you what I recognize the most important things in achieving that goal. The FBR cycle can not be commercialized without establishing both reactor and fuel cycle technologies. When FBRs are commercialized, it is easily found that the issues to be required in nuclear energy like nuclear non-proliferation,transparency and public acceptance must be more important than now. FBRs can not operate without fuel cycle activities including reprocessing, mixed oxide (MOX) fuel fabrication and MOX fuel transportation and so on. These activities require more effective and efficient safeguards, strengthened security measures in order to prevent proliferation and public acceptance than Light Water Reactors (LWR) operation because of the handling large amount of plutonium in the FBR cycle. Additionally, it should be noted that nuclear reactors are not possible to be installed and operated without agreement and cooperation of local residents. It is likely that more efforts are necessary to get public acceptance for the FBR cycle facilities with large amount of plutonium than LWRs. Treaty of the Non-Proliferation of Nuclear Weapons (NPT) requires that all the peaceful nuclear activities should be reported to the International Atomic Energy Agency (IAEA) and put under its safeguards. Japan

  19. Research and Development Policy on FBR Cycle Technology in Japan

    The fast breeder reactor (FBR) is a quite effective and realistic measure for establishing a long term, stable energy supply and for preventing global warming. In Japan, the FBR research and development project, named FBR Cycle Technology Development (FaCT), has been operational since April 2006. In this project, the combination of a sodium cooled fast reactor using oxide fuel and advanced aqueous reprocessing, as well as the simplified pelletizing fuel fabrication, is being developed principally as the most promising concept of FBR cycle technology to be commercialized, aiming at introducing the demonstration FBR by around 2025, and the commercial FBR before approximately 2050. Research and development for the establishment of the innovative technologies, which can meet design requirements for the demonstration FBR, has been steadily progressing. The adoption of the innovative technologies will be decided by judging their applicability and the conceptual designs of demonstration and commercial FBR cycle facilities by 2015. Consequently, the development of innovative technologies should be completed by 2015. Thereafter, the FaCT project will enter the introduction stage through a system demonstration. (author)

  20. FBR related test facilities data base

    The questionnaire of main specifications, test performance and features of each FBR related test facility in the O-arai Engineering Center were made from 2001 to 2002. This report equipped these questionnaires with database. Two tables list 134 facilities. These related test facilities contains the safety test, thermal hydraulics test, test facilities for structure, reactor, Na related test, irradiation rig, fuel monitoring facility and apparatus and others (failed fuel detection and location, helium accumulation fluence monitor measurement system, inductively coupled plasma mass spectrometer, laser resonance ionization mass spectrometry system, pressurized resistance welding equipment, fuel inspection system and inductively coupled plasma mass spectrometer). This report contains all questionnaires as data. (S.Y.)

  1. The Swiss contribution to the FBR development

    The program of fast breeder reactors development in Switzerland is considered from two points of view: energy self-sufficiency and optimization of fuel cycle. Research and development program covers: safety features of LMFBRs, development of mixed carbide fuel elements, study of steam generators transient behaviour, influence of various cooling concepts on thermal efficiency, techniques for detecting cover gas bubbles in the primary sodium circuit. This paper includes cost od the research and development activities as well as description of the future aims of the FBR projects

  2. Present status and future outlook of FBR development in Japan. History and present status of FBR development in Japan

    In 1966, the Japan Atomic Energy Commission drew up its 'Long-term Plan for Nuclear Energy Utilization and Development' which stated that the FBR should constitute the mainstream of nuclear power generation in the future. The plan defined a national project for the step-by-step development of an experimental reactor, a prototype reactor and a demonstration reactor aimed at commercializing the FBR by about 1990. At present, it is anticipated that the FBR will be commercialized during the first half of the 21st century. The 50 MWth experimental fast reactor 'Joyo' was taken critical for the first time in 1977. In 1981, the core of Joyo was changed to improve its irradiation capability as a test-bed for FBR fuels and materials, and since 1982 it has operated successfully in this role. It is planned to further upgrade the core of Joyo in 2002. The pre-operational tests of Monju - a prototype FBR power-generating plant designed to have an output of 280 MWe -, which was started in 1992, were interrupted by a sodium leakage accident in the secondary heat transport system in December 1995 during a 40 per cent power test. Monju remains shut down since the accident. Following the sodium leakage accident, the Japan Atomic Energy Commission concluded: - The FBR is a promising future source of non-fossil fuel energy, and it is appropriate to pursue FBR commercialization. - Monju should be utilized in order steadily to accumulate FBR data, and to acquire data for new R and D fields. - The development program for a commercial FBR should be flexible, giving priority to safety and economics and considering the future energy situation. (orig.)

  3. Design criteria for FBR core components

    This paper outlines the general approach adopted in France to take into account the specific behavior of irradiated steel for functional and structural verification of fast breeder reactor core components. Functional verification deals with the distortions which appear in structures as a result of void swelling and irradiation creep. Specific criteria must be defined to limit these distortions to acceptable values: these criteria are highly dependent on the subassembly and core design. Structural verification deals with modifications of the mechanical properties of steel submitted to FBR flux. Conventional standards and rules are not applicable, and a new methodology must be defined to take into account the new characteristics of irradiated steel. The general R and D program set up to investigate these areas is presented here as it is implemented in France but with emphasis on integration in a joint European program

  4. Evaluation of core distortion in FBR

    The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

  5. Technological innovations for FBR reactor cooling system

    The fast breeder reactor (FBR) is expected to be commercialized early in the 21st century. In order to realize this goal, technological innovations are desired in order to extensively enhance economic performance, and improvement of the reactor cooling system is of primary importance in this regard. Over the past 10 years, Toshiba has developed a succession of new technologies in the field of reactor cooling systems, including a compact type intermediate heat exchanger (IHX), an integral once-through type steam generator (SG), a double-wall-tube type steam generator, and a sodium-immersed high-temperature type electromagnetic pump (EMP). As a synthesis of the fruits of such research and development we have formulated innovative concepts for a reactor cooling system and its constituent components. These advances in research and development activities will significantly contribute to the commercialization of FBRs. (author)

  6. Principles of MONJU maintenance. Ageing phenomena of FBR for maintenance

    Because a fast breeder reactor (FBR) uses sodium as a coolant and there exist systems and components unique to FBR, aging phenomena to be considered in FBR are quite different from ones in a light water reactor (LWR). Therefore it is required to make a maintenance plan for FBR by considering the characteristics of FBR. In the Japanese LWRs, aging phenomena which occur in every part of all safety-related components have been extracted based on the O and M experiences and the technical evaluation on aging phenomena. In this study, aging phenomena or degradation mechanism which may occur in FBR was extracted based on the O and M experiences in the Japanese FBRs including the experiences of the overseas FBRs by using the three step method that is similar to one used in the Japanese LWRs. This paper reports the results of the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM), which consists of engineers from the electric utilities having LWRs, engineers from the plant vendors, and experienced people from the academic circles. (author)

  7. Research and development policy on FBR cycle technology in Japan

    The Fast Breeder Reactor (FBR) and its fuel cycle (hereinafter 'FBR cycle') technology will provide harmonic solutions for global energy resource and environment issues.In Japan, the significance of FBR cycle technology development has been recognized for decades. The Japan Atomic Energy Agency (JAEA) has been the principal agency for the FBR cycle development in Japan. The experimental fast reactor, Joyo, had been successfully operated for about 30 years, beginning in 1977. The prototype FBR, Monju, achieved initial criticality in 1994. Monju is designed on the basis of research results gleaned from Joyo. Monju has the role of confirming the technological data base for design and safety evaluation tools, and of accumulating operation experiences for sodium-cooled reactors, with an eye toward commercialization. Both reactors' operations have been suspended since 2007 and 1995, respectively, due to troubles; presently, JAEA is preparing for the re-launch of operations. Furthermore, the development of FBR spent fuel reprocessing technologies was initiated in 1975, and JAEA has successfully achieved MOX fuel fabrication at the Plutonium Fuel Center, as far back as 1972. In 1999, the 'Feasibility Study on Commercialized FBR Cycle Systems (FS)' was initiated to present an appropriate picture of FBR cycle technology commercialization by 2015, as well as its research and development (R and D) program. In this study, conceptual design features were evaluated in order to select promising FBR cycle systems that could meet the design requirements that embodied the five development targets: 1) safety; 2) economic competitiveness; 3) efficient utilization of nuclear fuel resources; 4) reduction of environmental burden; and 5) enhancement of nuclear non-proliferation. As a result, the combination of sodium-cooled FBR with oxide fuel, advanced aqueous reprocessing and simplified pelletizing fuel fabrication was selected as the most promising concept for the FBR cycle system

  8. Sodium cooling FBR type reactor plant

    In a sodium cooling FBR type reactor plant, a steam electrolysis type hydrogen forming device using high temperature steams as the starting material and a steam turbine power generator operated by high pressure/high temperature steams are disposed. During day time when a power demand is increased, a steam switch valve on the side of the hydrogen forming device is closed and the steam switch valve on the side of the power generator is opened to introduce substantially entire amount of steams to the steam turbine. During mid night when the power demand is decreased, the steam switch valve on the side of the power generator is closed and the steam switch valve on the side of the hydrogen forming device is opened to introduce substantially entire amount of steams to the steam electrolysis device, or the exhausted gases from the steam turbine type power generator is heated again by a heat exchanger and low pressure/high temperature steams are introduced to the steam electrolysis device. In this case, electric power is applied between a hydrogen electrode and an oxygen electrode to form a hydrogen gas and an oxygen gas, and the hydrogen gas is stored in a hydrogen storage vessel. It can easily cope with the fluctuation of the power demand, as well as hydrogen can be efficiently produced. (N.H.)

  9. Improvement of FBR multi-sided comparison system

    In this study, the general assessment system developed last year was improved so that we could perform multi-sided comparison and evaluation between FBR cycle reactor models quantitatively and objectively. At first, we developed multi-sided evaluation system written in VBA. And then many processes (calculation of coefficients for utility functions, analysis by MUF, and pair comparison, etc.) could be performed on a Personal Computer with simple operation. And the following functions were also added in the system; consideration of uncertainty in design data on evaluated results, estimation of the potential of FBR, the absolute evaluation method to get over the weakness of AHP, and so on. Moreover, a concept of comparison with compact reactors and the way of including a new evaluation item, public acceptance, in the evaluation structure are discussed. Although it is important to consider public acceptance for introduction of FBR, it was not included in the assessment structure of the previous study. In this study, we pointed out that public acceptance as a new assessment item of the FBR cycles evaluation system should be included. Ideas to include the item in the system in phase 2 were also discussed. Due to the improvement of the system in this study, we believe that the FBR cycle's evaluation system is becoming more sophisticated than the previous version. (author)

  10. FBR research and development at Oarai Engineering Center

    Construction and operation of the 'JOYO' and accumulation and filling up of the Advanced Technology were improved at the Oarai Engineering Center from 1970. These results have been reflected to construction and operation of the FBR 'MONJU'. And we tackle on investigation of the cause of the sodium leak accident happened at the 'MONJU'. In future, technology will be carried to extremes to be able to make the Commercialization till 2030, in line with the 'Long-term FBR R and D and Utilization Programs'. And technology systems will be composed to correspond to many kinds of needs in the Commercialization age. This report introduces R and D present conditions and future developments of 'Safety', 'Core and Fuel' and 'Systems and components' that is the cores of the FBR peculiar technology. (author). 76 refs

  11. Development of FBR cycle data base system (II)

    In the 'Feasibility Study on Commercialized FBR Cycle Systems (F/S)', scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show the significance of the FBR cycle system introduction concretely are performed while design studies for FBR plants, reprocessing systems and fabrication systems are conducted. In these evaluations, future society of various conditions and situation is assumed, and investigation and analysis about needs and social effects of FBR cycle are carried out. In this study, promising FBR cycle concepts are suggested by taking information such as domestic and foreign policies and bills, an economic prediction, a supply and demand prediction of resources, a project of technology development into consideration in addition to system design information. The development of the FBR Cycle Database which this report introduced started in 1999 fiscal year to enable managed unitarity and searched reference information to use for the above scenario evaluations, cost-benefit evaluations and system characteristic evaluations. In 2000 fiscal year, its prototype was made and used tentatively, and we extracted the problems in operation and functions from that, and, in 2001 fiscal year, the entry system and the search system using the Web page were made in order to solve problems of the prototype, and started use in our group. Moreover, in 2002 fiscal year, we expanded and improved the search system and promoted the efficiency of management work, and use in JNC through intranet of the database was started. In addition, as a result of having made the entry of about 350 data in 2002 fiscal year, the collected number of the database reaches about 7,250 by the end of March, 2003. We are to continue the entry of related information of various evaluations in F/S phase 2 from now on. In addition, we are to examine improvement of convenience of the search system and cooperation with the economy database. (author)

  12. Myristylation of FBR v-fos dictates the differentiation pathways in malignant osteosarcoma

    1996-01-01

    Myristylation of FBR v-fos, a c-fos retroviral homologue that causes osteosarcomas in mice, determines many of its transcriptional properties in vitro. To determine whether myristylation of FBR v-fos contributes to in vivo tumorigenicity, we examined its transforming capability in comparison to a nonmyristylated FBR v-fos (G2A-R). Retroviral infections with FBR v-fos and G2A-R transform BALB/c-3T3 cells. The number, size, and cellular morphology of foci generated by both FBR and G2A-R are ind...

  13. Core disruptive accident analysis using ASTERIA-FBR

    Japan Nuclear Energy Safety Organization (JNES) is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents (CDA) of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. It was found that Dynamic-GMVP is confirmed to be basically applicable to the CDA phenomena. It was found that, however, applying GMVP to the CDA calculation is less reasonable than PARTISN since the calculation load of GMVP is too large to meet the required calculation accuracy, although the Monte-Carlo method is based on the actual neutron behavior without any discretization of space and energy. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy

  14. Shaking table test on base isolated FBR plant model, 1

    Shaking table test on seismically isolated FBR plant model is carried out. The model is three story steel frame structure supported by nine laminated rubber bearings which are reduced to a scale of 1/15. The effectiveness of base isolation system is verified through this study. (author)

  15. Development of the demonstration FBR and its commercialization

    The FBR being developed in Japan is expected to operate commercially by the year 2030 or so. The construction of the first demonstration FBR ''DFBR-1'', which follows the experimental FBR ''JOYO'' and the prototype FBR ''MONJU'' is scheduled for early 2000s. The Atomic Energy Commission adopted these policies in the new version of ''The Long Term Program for Research, Development and Utilization of Nuclear Energy''. The Japan Atomic Power Company has progressed design studies of the DFBR-1 and R and Ds. Electric utilities settled the major design specifications of the DFBR-1 such as the top entry loop type as the reactor type and about 660 MWe of plant capacity. To achieve the goals for the commercialization of FBRs, such as a high level of safety and reliability comparable with a LWR and further improvement of economy, many innovative technologies and system improvements will be required in commercial FBRs. This paper describes the present status of development of the DFBR-1. (author)

  16. Development of the in-vessel repairing technology with friction stir welding method for FBR

    Application of Friction Stir Welding (FSW) method to in-vessel repairing of FBR has been studied. Material and shapes for the FSW tool were surveyed from the viewpoint of the in-vessel repairing of the reactor vessel and the in-vessel components made of stainless steel. It was also found that load control method was more preferable rather than position control method for the in-vessel repairing machine. Repair on SS316L austenitic stainless steel plates with a slit-type artificial crack has successfully demonstrated in both argon gas environment and in liquid sodium. Plunging downforce for FSW was not more than 30 kN, which is within realistic force capability of the in-vessel repairing machine. The welding speed had to be much slower than that in the argon environment to input enough friction heat for compensating the heat removed to the liquid sodium. Tensile strength of the repaired specimens was the same level as the base metal after a heat treatment, which corresponds to the 40 years of FBR operation. Weld bead and HAZ, which are more likely to have defects compared to the base metal, were also FSWed in the argon environment and it was found that the optimum process conditions were the same as the base metal. Considering the repairing on the curved in-vessel components, FSW on the tilted plate and curved specimen have been performed. Allowable tilt angle has been studied and corresponding length on the in-vessel component has been estimated. A concept of the in-vessel repairing machine has been established. The machine is inserted through the ISI hole to the inside of the reactor vessel and reaches to the place to be repaired with jointed-robotic arm. (author)

  17. Correlations among FBR core characteristics for various fuel compositions

    In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations. (author)

  18. Development on FBR fuel reprocessing technology in PNC

    Spent nuclear fuel reprocessing is positioned as a key part of plutonium recycling system based on fast breeder reactor (FBR). Reprocessing technology for FBR spent fuel has been developed to complete the nuclear fuel cycle for the past two decades by the Power Reactor and Nuclear Fuel Development Corporation (PNC). FBR spent fuel has a couple of unique characteristics when comparing with light water reactor (LWR) spent fuel such as adoption of wrapper tube on fuel assembly, high plutonium content, high burn up and so on. Its main process is based on purex-type LWR reprocessing plant, which is being demonstrated by Tokai reprocessing plant in Japan. Application of a new technology has been tried instead of a conventional one. Based on this background, R and D has been performed vigorously to develop a new type of equipment, such as disassembly machine, dissolver, clarifier and extractor, and to construct advanced processes such as salt free process or rack modulated remote maintenance system. In consequence of R and D work, prototype equipment which forecast practical use are developed and their performance confirmed at a new facility named Recycling Equipment Test Facility (RETF) which is now under construction

  19. Preparation for Korean-French joint study on FBR introduction to Korea

    A joint study between Korea and France for a period of 2 years (1983-1984) has been planned to extract the prerequisite infrastructure and to establish an optimum system for preparation of the introduction of FBR to Korea, for an FBR plant is expected to be commercialized around the turn of this country. Prior to the joint study, we examined in this work the scope of work to be performed by each partner, through the analyses of reactor technology, international relationship and international economic trends. In the reactor technology, we examined the following items; plutonium availability, differences in fuel fabrication, reprocessing between oxide fuel and mixed oxide fuel, sodium technology, nuclear and thermal characteristics of FBR mixed oxide core, licensibility of pool type FBR in the light of U.S. safety regulatory standards, unique French experiences and technology which cannot be acquired from other FBR developing countries. Also, manpower needed in FBR project was estimated. In the international economy, we examined the following items; prospects of the international economy in 1990's, world economic trends and the Korean position on them, the effect of the introduction of FBR into Korea, etc. Through the above analyses, it was found that FBR industry would be an effective strategic one in case that FBR would be introduced early by the enhancement of national investment capability and multilateral foreign trades. (Author)

  20. FBR Metallic Materials Test Manual(English Version) (Manual)

    小高 進; 加藤 章一; 吉田 英一

    2003-01-01

    For the development of the fast breeder reactor, this manual describes the method of in-air and in-sodium material tests and the method of organization the data. This Previous manual has revised in accordance with the revision of Japanese Industrial Standard (JIS)and the conversion to the international unit. The test methods of domestic committees such as the VAMAS (Versailles Prject on Advanced Materials and Standards) workshop were also referred. The material test technologies accumula...

  1. Uranium Recovery in LWR Reprocessing and Plutonium/Residual Uranium Conditioning in FBR Reprocessing for the Transition from LWR to FBR

    In order to flexibly manage the transition period from LWR (light water reactor) to FBR (fast breeder reactor), the authors investigated the transition scenario and proposed the Flexible Fuel Cycle Initiative (FFCI). In FFCI, LWR spent fuel reprocessing only carries out the removal of about 90% uranium that will be purified and utilized in LWR after re-enrichment. The residual material (∼40% U, ∼15% Pu and ∼45% other nuclides) is transferred to temporary storage and/or FBR spent fuel reprocessing to recover Pu/U followed by FBR fresh fuel fabrication depending on the FBR introduction status. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR spent fuels, that is smaller LWR reprocessing facility, spent LWR fuel reduction, storage and supply of high proliferation resistant and high Pu density material that can flexibly respond to FBR introduction rate changes. The Pu balance was calculated under several cases, which revealed that the FFCI could supply enough Pu to FBR in any cases. (authors)

  2. Uranium Recovery in LWR Reprocessing and Plutonium/Residual Uranium Conditioning in FBR Reprocessing for the Transition from LWR to FBR

    Tetsuo Fukasawa; Junichi Yamashita; Kuniyoshi Hoshino [Hitachi-GE Nuclear Energy, Ltd., 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Akira Sasahira [Hitachi, Ltd. (Japan); Tadashi Inoue [Central Research Institute of Electric Power Industry (Japan); Kazuo Minato [Japan Atomic Energy Agency (Japan); Seichi Sato [Hokkaido University (Japan)

    2008-07-01

    In order to flexibly manage the transition period from LWR (light water reactor) to FBR (fast breeder reactor), the authors investigated the transition scenario and proposed the Flexible Fuel Cycle Initiative (FFCI). In FFCI, LWR spent fuel reprocessing only carries out the removal of about 90% uranium that will be purified and utilized in LWR after re-enrichment. The residual material ({approx}40% U, {approx}15% Pu and {approx}45% other nuclides) is transferred to temporary storage and/or FBR spent fuel reprocessing to recover Pu/U followed by FBR fresh fuel fabrication depending on the FBR introduction status. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR spent fuels, that is smaller LWR reprocessing facility, spent LWR fuel reduction, storage and supply of high proliferation resistant and high Pu density material that can flexibly respond to FBR introduction rate changes. The Pu balance was calculated under several cases, which revealed that the FFCI could supply enough Pu to FBR in any cases. (authors)

  3. Effect of pressure on the free ion and crystal field parameters of Sm2+ in BaFBr and SrFBr hosts

    The emission spectra of Sm2+ doped in BaFBr and SrFBr hosts were measured at 10 K from ambient pressure to 8 GPa. The crystal field energy levels determined from the emission spectra were used to extract the free ion parameters (Fk and ζ) and crystal field parameters (Bqk). The variation of Fk and ζ as a function of pressure was studied systematically and was discussed in relation to the central field and symmetry restricted covalency models. The change of the spin–orbit coupling parameter (ζ) with pressure for SrFBr:Sm2+ showed very different behavior than in other matlockite hosts. Moreover the variation of Bqk under pressure was studied. The pressure dependence of the Bqk was described quantitatively using the Superposition Model (SM) with the help of structural parameters as a function of pressure, obtained from periodic DFT calculations. The validity of the SM was tested for Sm2+ in BaFBr and SrFBr. It is shown that this model does not apply to SrFBr, in contrast to other matlockite host materials. - Highlights: ► Luminescence at 10 K from ambient pressure to 8 GPa were measured for BaFBr:Sm2+ and SrFBr:Sm2+. ► The free ion (F2 and ζ) and crystal field parameters were determined as a function of pressure. ► Crystal field strength shows different trend for the two compounds. ► The superposition model is unable to explain the unusual behavior of intrinsic parameters in SrFBr:Sm2+.

  4. Analysis of passive shutdown capability for a loss of flow accident in a medium sized liquid metal fast breeder reactor

    The passive shutdown capability of a medium sized (500 MWe) liquid-metal fast breeder reactor with oxide, carbide and metal fuels has been analysed for a loss of flow accident using static and dynamic analysis methods. The carbide fuel is assumed to be He-bonded as well as Na-bonded. The dependence of the passive safety on the flow halving time constant of the loss of flow incident and the feedback components, like radial core expansion due to subassembly spacer pad heating and differential control rod expansion due to heating of the control rod suspension mechanism, is highlighted. (author)

  5. Description of JNC's analytical method and its performance for FBR cores

    The description of JNC's analytical method and its performance for FBR cores includes: an outline of JNC's Analytical System Compared with ERANOS; a standard data base for FBR Nuclear Design in JNC; JUPITER Critical Experiment; details of Analytical Method and Its Effects on JUPITER; performance of JNC Analytical System (effective multiplication factor keff, control rod worth, and sodium void reactivity); design accuracy of a 600 MWe-class FBR Core. JNC developed a consistent analytical system for FBR core evaluation, based on JENDL library, f-table method, and three dimensional diffusion/transport theory, which includes comprehensive sensitivity tools to improve the prediction accuracy of core parameters. JNC system was verified by analysis of JUPITER critical experiment, and other facilities. Its performance can be judged quite satisfactory for FBR-core design work, though there is room for further improvement, such as more detailed treatment of cross-section resonance regions

  6. An overview of DHR studies for Indian FBR program

    Indian Fast Reactor Program is entering into an important phase. Fast Breeder Test Reactor (FBTR) at Kalpakkam, after it was made critical with a small core in October 1985, was operated at low power levels until the mid 1992. Various tests were performed and the tests were completed with the demonstration that thermal power up to 1 MWt could be safely removed solely by natural convection in the steam generator casing air flow path. Theoretical studies performed using simplified thermal and hydraulic models of the FBTR reactor system satisfactorily predicted the natural circulation heat removal capability of the reactor system. It was indeed predicted that natural convection in primary and secondary sodium circuit could also be depended upon for heat removal while keeping all the temperatures within limits. 1 MWt of thermal power being the short term post shutdown fission product decay power level for a full 40 MWth core, the above test has convincingly demonstrated that natural convection mode of decay heat removal can be relied upon. Further theoretical studies has shown that, with a primary flow halving time of 17s, even an unprotected (or without reactor trip) total power failure situation from full power operation can be sustained with reactor power reducing gradually to decay power levels because of the negative reactivity feedback and with all the temperatures well within the limits. This paper brings out the details of the tests performed and the results of the various theoretical predictions. Presently FBTR is poised to be operated up to a power level of 10 MWt. Simultaneously preliminary design of Prototype Fast Breeder Reactor (PFBR), a 500 MWe, 1200 MWt LMFBR has been completed. Presently it is in a detailed design phase. In this overview, the various options available for post shutdown decay heat removal in PFBR, the argument for selection of the normal heat transport path and Safety Grade Decay Heat Removal System (SGDHRS) as two diverse paths of heat

  7. Training report of the FBR cycle training facility in 2001

    The FBR Cycle Training Facility has been operating since Sep. 2000 for staff of FBR 'Monju' and ATR 'Fugen' and fireman of the local community, etc. In 2000, 6 courses of the sodium handling training were held 22 times and also eight courses of the maintenance training were 11 times and total participants were 305. In order to grade up the trainigs for providing restarting of 'Monju', the corresponding training course on sodium piping leakage accident was added as newly sodium handling training course and partial contents of some courses were devised based on some comments by the training discussing committee organized in our center. In 2001, seven sodium handling training courses were carried out 25 times and eight maintenance courses were conducted 11 times and total participated number was 220, i.e. 157 trainees for sodium training courses and 63 trainees for maintenance training courses. Additionally, since a new licensed sodium training course for sodium handling workers included head of worker will be introduced from next year owing to the fire accident of maintenance building in the experimental fast reactor 'JOYO' occurred Oct in 2001, which was coursed by sodium spontaneous combustion due to inefficient handling on tidy up of the sodium handling working, training curriculum, training text and questions for examination are provided as for preparing its course cooperated with Oarai Engineering Center. (author)

  8. A final report on stress analysis of seal disc of 500 MWe PHWR

    In Pressurised Heavy Water Reactor (PHWR) on-power refuelling is done by use of fuelling machine. Before refuelling, sealing plug assembly is removed from the end-fitting of the coolant channel and after refuelling the sealing plug is reinstalled back in to the end-fitting. The seal disc is a part of sealing plug assembly. Its function is to create sealing action for the heavy water inside the coolant channel. A systematic developmental work is done to arrive at a final configuration of the seal disc. This is done to minimise the stresses in the body of the seal disc and at the same time to obtain required seating reaction to avoid heavy water leakage. It is observed that stresses computed for the final configuration by linear elastic analysis are more than the allowable value as per ASME Section III, Division 1. This calls for elasto-plastic analysis to find out collapse load to satisfy ASME codal limits as per special provision of NB-3228.1 (1986). The elasto-plastic analysis showed that the seal disc meets ASME codal limits for all stages of loading. (author). 8 refs., 2 tabs., 7 figs

  9. Synergy of automation and human action in fuel handling control system of 500 MWe PHWR

    On-line refueling process is carried out in PHWRs on a routine basis. Each refueling cycle involves a large number of sequential operations to be executed safely and satisfactorily and hence is time consuming. Operations on the system are performed as per the detailed prescribed procedures. This necessitates introduction of automation of these operations; however the extent of automation has to be based on consideration of professional motivation and psychological well-being of the operator. This paper describes the nature of fuel handling operations and how man and machine work together in a complementary manner to achieve the objectives. (author). 2 refs., 7 figs., 1 tab

  10. Nuclear heat generation race on the walls and floor of calandria vault of 500 MWe PHWR

    Nuclear heat generation rate profile along the walls and floor of calandria vault is required to analyse its strength during the life time of reactor power operation and to ascertain adequacy of cooling and shielding arrangements. Hence detailed heat generation profile on the face and across the thickness of walls and floor is generated. (author). 7 refs., 1 fig., 3 tabs

  11. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  12. CONCEPTUAL DESIGN AND ECONOMICS OF A NOMINAL 500 MWe SECOND-GENERATION PFB COMBUSTION PLANT

    A. Robertson; H. Goldstein; D. Horazak; R. Newby

    2003-09-01

    Research has been conducted under United States Department of Energy Contract DE-AC21-86MC21023 to develop a new type of coal-fired plant for electric power generation. This new type of plant, called a Second Generation Pressurized Fluidized Bed Combustion Plant (2nd Gen PFB), offers the promise of efficiencies greater than 48 percent, with both emissions and a cost of electricity that are significantly lower than those of conventional pulverized coal-fired (PC) plants with wet flue gas desulfurization. The 2nd Gen PFB plant incorporates the partial gasification of coal in a carbonizer, the combustion of carbonizer char in a pressurized circulating fluidized bed boiler, and the combustion of carbonizer syngas in a gas turbine combustor to achieve gas turbine inlet temperatures of 2300 F and higher. A conceptual design and an economic analysis was previously prepared for this plant. When operating with a Siemens Westinghouse W501F gas turbine, a 2400psig/1000 F/1000 F/2-1/2 in. Hg. steam turbine, and projected carbonizer, PCFB, and topping combustor performance data, the plant generated 496 MWe of power with an efficiency of 44.9 percent (coal higher heating value basis) and a cost of electricity 22 percent less than a comparable PC plant. The key components of this new type of plant have been successfully tested at the pilot plant stage and their performance has been found to be better than previously assumed. As a result, the referenced conceptual design has been updated herein to reflect more accurate performance predictions together with the use of the more advanced Siemens Westinghouse W501G gas turbine. The use of this advanced gas turbine, together with a conventional 2400 psig/1050 F/1050 F/2-1/2 in. Hg. steam turbine increases the plant efficiency to 48.2 percent and yields a total plant cost of $1,079/KW (January 2002 dollars). The cost of electricity is 40.7 mills/kWh, a value 12 percent less than a comparable PC plant.

  13. Dynamic analysis of containment building of 500 MWe pressurised heavy water reactor using finite element method

    The reactor building essentially comprises of a double containment system. The two containments viz. inner and outer containments (ICW and OCW), are axi-symmetric structures along with the raft and therefore an axi-symmetric finite element analysis of these should suffice. However, the inner containment is connected with the reactor internals (i.e. internal structures and calandria vault) at EL-130.0 meter elevation and thereby will have an interaction with the reactor internals. Exact modelling of this interaction effect is a formidable task since the reactor internals are not axi-symmetric. Hence, an equivalent axi-symmetric model of the reactor internals was evolved in such a way that the dynamic characteristics of the interaction effect are preserved. Analysis of the containment building has been carried out for two mutually perpendicular horizontal directions (N-S) and E-W) and the vertical direction using response spectrum and time history technique. Due credence was given to soil-structure interaction. This report presents the results and conclusions arrived at for these analyses. (author). 18 refs., 81 figs., 68 tabs., 3 appendixes

  14. Development of liquid poison injection system (SDS-2) for 500 MWe PHWRs

    A secondary shut-down system (SDS-2) in the form of a mecahnism for introducing poison into the moderator of the PHWR is under development in Reactor Engineering Division of BARC. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator. This report details the development work being done on this poison injection system. An experimental facility was set up to measure the poison jet growth rate and the jet spread after injection, and mathematical models were developed to convert the observed jets into reactivity worth values. A description of the work and the computed results are presented. (author). 21 graphs. , 15 tabs

  15. Fuel Behavior Simulation Code FEMAXI-FBR Development for SFR Core Disruptive Accident Analysis

    Japan Nuclear Energy Safety Organization (JNES) has been developing ASTERIA-FBR code system for SFR core disruptive accident analysis to contribute as a part of the regulation activity for Japanese prototype FBR, MONJU. The ASTERIA-FBR code system consists of detailed fuel behavior analysis module (FEMAXI-FBR), neutronic Monte-Carlo calculation module (GMVP), and thermal hydraulic module (CONCORD). The calculation scope of the ASTERIA-FBR covers the initiating, transitional and post disassembly expansion processes. The FEMAXI-FBR is based on LWR fuel behavior simulation code FEMAXI-6 and modified the material properties and the calculation models under steady state and transient operational condition. The FEMAXI-FBR has been verified in steady state calculations compared with those of SAS-4A code. Furthermore, the code has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient calculations. Through these verification and validation, good agreement has been obtained with the FP-gas release ratio, the fuel restructuring, the gap width between pellet and cladding, and the fuel pin failure position. (author)

  16. Present state of development of demonstration FBR and prospect of practical use

    As for the FBR development in Japan, the Atomic Energy Commission revised the long term plan on the research, development and utilization of atomic energy in June, 1994, and under the basic policy that through the considerable period of using LWRs together, FBRs will be adopted as the main nuclear power plants in future, it was decided to establish FBR technology system so that the practical use of FBRs becomes feasible by about 2030 through two demonstration FBRs following the experimental FBR 'Joyo' and the prototype FBR 'Monju'. The Monju started power generation and transmission in August, 1995, but secondary sodium leak accident occurred in December, 1995, and at present it is stopped. The demonstration FBR No. 1 is a top entry type loop reactor, and the power output is about 660 MWe. The start of construction is scheduled at the beginning of 2000s. The research on the whole plant design is carried out as the research on the optimization of demonstration FBR plant for three years from fiscal year 1994. The design of the demonstration FBR No. 1, the research and development for it, the prospect of the practical use and the research and development for the practical use are reported. (K.I.)

  17. Development of the tool for generating ORIGEN2 library based on JENDL-3.2 for FBR

    ORIGEN2 is one of the most widely-used burnup analysis code in the world. This code has one-grouped cross section libraries compiled for various types of reactors. However, these libraries have some problems. One is that these libraries were developed from old nuclear data libraries (ENDF/B-IV,V) and the other is that core and fuel designs from which these libraries are generated do not match the current analysis. In order to solve the problems, analysis tool is developed for generating ORIGEN2 library from JENDL-3.2 considering multi-energy neutron spectrum. And eight new libraries are prepared using this tool for analysis of sodium-cooled FBR. These new libraries are prepared for eight kinds of cores in total. Seven of them are made by changing core size (small core - large core), fuel type (oxide, nitride, metal) and Pu vector as a parameter. The eighth one is a Pu burner core. Burnup calculation using both new and original libraries, shows large difference in buildup or depletion numbers of nuclides among the libraries. It is estimated that the analysis result is greatly influenced by the neutron spectrum which is used in collapse of cross section. By using this tool or new libraries, it seems to improve evaluation accuracy of buildup or depletion numbers of nuclides in transmutation research on FBR fuel cycle. (author)

  18. Current status of development of Demonstration Fast Breeder Reactor and prospect of FBR commercialization

    The Demonstration Fast Breeder Reactor (DFBR) is the next step of FBR development following the prototype fast breeder reactor 'MONJU'. The DFBR is now under development by The Japan Atomic Power Co. (JAPC) under the sponsorship of 9 Japanese electric power companies and Electric Power Development Co., Ltd. The JAPC has been performing the design study and R and D for DFBR in cooperation with Power Reactor and Nuclear Fuel Development Corp. (PNC), Central Research Institute of Electric Power Industry (CRIEPI) and Japan Atomic Energy Research Institute (JAERI). This report describes the prospect of FBR commercialization and the current status of new technology for DFBR and innovative technology FBR commercialization. (author)

  19. Development of ISI system for FBR MONJU reactor vessel

    This paper describes the development of a new inspection robot and support system used for the in-service inspection (ISI) of the reactor vessel of the FBR MONJU. The inspection is carried out using a CCD camera for visual tests and an EMAT for volumetric tests at elevated temperature (200 degree C) and high irradiation dose condition (10 Sv/hr). The applicability of a commercial CCD camera and an EMAT have been confirmed by experiments under severe conditions. In addition, reduction of the robot weight from 47 kg to 34 kg is achieved by optimization of all parts in the robot, these results in a reduction of wear and replacement frequency of tires. A performance improvement with EMAT was attained by a new magnet arrangement and a signal treatment method. (author)

  20. Control rod drives for use in FBR reactors

    Purpose: To improve thermal fatigue-resistance and corrosion resistance thereby increasing the working life of control rod drives. Constitution: Control rod drives for use in FBR type reactors are made of invar alloy substrate and ceramic layers composed of cordierite composition coated on the surface of the substrates. Since the thermal expansion coefficients for the invar alloy and the ceramic layer are of about 10-6 - 10-7 respectively, thermal stresses due to the transient thermal changes in the surface are small to thereby prevent the generation of fatigue cracks, particularly, those fatigue cracks in the circumferential direction at the surface. Accordingly, blocking of coolants due to the stripping of the cracked portions can be prevented. (Moriyama, K.)

  1. Development of visual inspection technique under sodium in FBR

    The reactor vessel of a fast breeder reactor (FBR) is filled with optically opaque liquid sodium. Therefore, the ultrasonic imaging technique is useful for inspecting in-vessel structures in sodium. We have developed a high-speed and high-resolution three-dimensional image processing technique. For imaging in the sodium, a two-dimensional matrix transducer and the M-series transmitting signal were used. Cross-correlation processing between the transmitted signal and received signal was used to enhance the S/N ratio. Image synthesis also attempts enhancement of resolution by means of the synthetic aperture focusing (SAFT). It has been confirmed that clear and high-resolution three-dimensional ultrasonic images are acquired at a distance of 0.8m in the in-water visualizing test. (author)

  2. A sociotechnical analysis of the French FBR programme: Evaluation as a cornerstone

    Based on a three-year study of the French FBR program, this paper aims at showing the dynamics of evaluation of FBR demonstrators, with methodological inputs from the “Science and Technology Studies” branch of sociology. Such a reactor has to demonstrate the feasibility - including safety, technical and economic viability - of a promising technology regarded as a potentially inexhaustible energy source for the future. The research shows that until the mid-seventies, the need for an FBR fleet was regarded as urgent, entailing a focus on proving technical feasibility with demonstration reactors. But after the mid-seventies, the evaluation of FBR projects gave more importance to other topics: not only did they have to prove the technical feasibility of the programme, but also its safety and economic viability. The analysis of the Superphénix case is used to illustrate the difficulty of reconciling the three elements of the evaluation in a changing context. (author)

  3. The present status of development of high chromium steel for FBR

    Authors perform a series of material tests for some high Cr steels to propose the most suitable high Cr steel specification for FBR pipes. Firstly, thermal expansion and heat conductivity of several high Cr steels are measured to predict optimum Cr content for FBR structural material. Secondly, influence of heat treatment conditions on long term ductility and toughness is studied to obtain the suitable properties for FBR components. Thirdly, focusing on tungsten (W) and molybdenum (Mo) which may form Laves phase, the most optimum balance of these elements is investigated based on long-term material tests and metallurgical examination results. Considering the results of the studies, a provisional specification of high Cr steel for FBR pipes is proposed. (orig.)

  4. Continuous Ethanol Production Using Immobilized-Cell/Enzyme Biocatalysts in Fluidized-Bed Bioreactor (FBR)

    Nghiem, NP

    2003-11-16

    The immobilized-cell fluidized-bed bioreactor (FBR) was developed at Oak Ridge National Laboratory (ORNL). Previous studies at ORNL using immobilized Zymomonas mobilis in FBR at both laboratory and demonstration scale (4-in-ID by 20-ft-tall) have shown that the system was more than 50 times as productive as industrial benchmarks (batch and fed-batch free cell fermentations for ethanol production from glucose). Economic analysis showed that a continuous process employing the FBR technology to produce ethanol from corn-derived glucose would offer savings of three to six cents per gallon of ethanol compared to a typical batch process. The application of the FBR technology for ethanol production was extended to investigate more complex feedstocks, which included starch and lignocellulosic-derived mixed sugars. Economic analysis and mathematical modeling of the reactor were included in the investigation. This report summarizes the results of these extensive studies.

  5. Development of weld plugging for steam generator tubes of FBR

    This study was undertaken to develop a method of weld plugging of the heat-exchanger tubes of steam generator of Prototype FBR 'MONJU' in case these tubes are damaged for some reason. We studied mainly the shape of plug, welding procedure and effect of postweld heat treatment (PWHT). Evaporator tube sheet, tube and plug are made of 2-1/4Cr-1Mo steel and usually preheating and PWHT will be required for welding of this steel. The results of this study is as follows. 1) Plug was designed to make butt joint welding with grooved tube sheet around the tube hole to satisfy the requirements of plug designing, stress analysis, and good weldability. 2) TIG welding process was selected and certified its good weldability and good performance. 3) PWHT can be done by using high frequency induction heating method locally and also designing the plug to weld joint with tube sheet which was grooved around the tube hole. 4) Mock up test was done and it was certified that this plugging procedure has good weldability and good performance ability for Non Destructive Inspection. (author)

  6. FPs release behavior from irradiated FBR MOX fuel

    Fission products (FPs) release behavior in the experiments (FP-1, FP-2) performed at JNC were evaluated from the viewpoint of source-term evaluation of FBR. Release fraction of FPs (Cs, Sb, Ru, Eu) in the experiments were calculated by NUREG-0772 model, and grain (hypothetical spheres) radius of fuel pellets and diffusion coefficient of FPs in grains for Booth model were calculated in the evaluation. Furthermore, the calculated values were compared with the values obtained in other experiments. The results of evaluation are summarized bellow. (1) The experiments (FP-1, FP-2) were performed in inert gas (Ar) condition, but FPs release fractions and diffusion coefficients in fuel grains almost agree with the values obtained in the other experiments which were performed in oxidized conditions. FPs release behavior in reduced condition containing such as sodium vapor are expected to be different from the behavior in FP-1 and FP-2. Some other experiments in reduced condition should be performed to investigate FPs release behavior in reduced conditions. Geometric grain radius obtained from metallographic images in used for the base data of diffusion coefficients evaluation, but the radius calculated from BET surface area is desirable. (2) Another release mechanism such as evaporation from the fuel pellet surface should be considered for more detailed evaluation of Cs composite and low volatile Ru and Eu release behavior, but, FPs chemical composition in fuel pellet, oxygen potential in carrier gas and FPs chemical composition in carrier gas, etc, should be evaluated in advance. (author)

  7. Principles of MONJU maintenance. Characteristic of MONJU maintenance and reflection of LWR maintenance experience to FBR

    A sodium cooled fast breeder reactor (FBR) has unique systems and components and different degradation mechanism from light water reactor (LWR) so that need to establish maintenance technology in accordance with its features. The examination of the FBR maintenance technology is carried out in the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM). As a result of the study such as extraction of Monju maintenance feature, maintenance technology benchmark between Monju and LWR components and survey of LWR maintenance experience, it is clear that principles of maintenance are same as LWR, necessity of LWR maintenance experience reflection and points to be considered in Monju maintenance. The road map to establish a FBR maintenance technology in the technical aspect became clear and it is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance. (author)

  8. Accelerating fissile fuel breeding in FBR with natural safety features

    To guarantee the rapid growth of the Chinese economy in 21st century, nuclear energy should be fully exploited, together with other renewable energies to replace coal and other depletive fossil fuels. Unfortunately, the major Chinese nuclear power plants under construction are mostly PWRs that would consume a lot of natural uranium during their operations. The availability of cheap natural uranium could seriously constraint the Chinese nuclear power development, unless artificial fissile fuel-plutonium is supplied from fast breeder reactors. The fissile nuclei production rate in the traditional fast breeders, however, seems too slow to match the rapid growth of nuclear power. A concept of the advanced fast breeder (AFBR) is introduced, therefore, to greatly accelerating the fissile fuel breeding process. In said breeder, the spherical hollow fuel elements and on-line refueling are adopted. The coolant flows horizontally through the annual fuel bed and bring out the full-fission power with natural circulation generated with the temperature difference between the inlet and outlet coolant flow. In such case, the fuel breeding rate could be greatly increased as the specific thermal power of the fuel inside the reactor core. The said AFBR is an improvement to the Russia-designed BREST FBR. It could transport, without the lead pump and high-elevation lead pool, all the fission heat solely via natural circulation under full power, as well as the decay heat after reactor shut-down, working just as a nuclear hot spring. No radioactivity would be released to the environments under any outside disasters. (authors)

  9. Evaluation of generation and heat transfer performance of low temperature AMTEC for sodium cooled FBR plant

    Design study of AMTEC system for Sodium Cooled FBR was performed. Heat power of the FBR reactor is 395 MWt, and the AMTEC system is used not only as electric power generator, but heat transfer device from primary coolant circuit to steam generation system. Evaluated maximum electricity generation efficiency was 49% which is increased about 9% higher than the case only steam generation system was used. Evaluated heat conduction area for AMTEC was about 1,000 m2. In the future, high performance AMTEC cell should be manufactured and design study of heat transfer equipment which has AMTEC system should be performed. (author)

  10. Method for selecting FBR development strategies in the presence of uncertainty

    This report describes the methods used to probabilistically analyze data related to the uranium supply the FBR's competitive dates, development strategies' time and costs, and economic benefits. It also describes the econometric methods used to calculate the economic risks of mistiming the development. Seven strategies for developing the FBR are analyzed. The various measures of a strategy's performance - timing, costs, benefits, and risks - are combined into several criteria which are used to evaluate the seven strategies. Methods are described for selecting a strategy based on a number of alternative criteria

  11. Development of FBR plant design program (FR-Design) for education

    Comprehensive understanding of FBR plants is required to realize safe and well-balanced plant design. Difficulties are related to their complex systems and multi-physics phenomena, which have mutual interaction among various professional fields. To overcome these problems, a total simulation program was developed to evaluate plant characteristics under steady state conditions and to survey allowable specifications among trade-off parameters. This program was proved to be effective for better understanding of FBR plants through unitization for course exercises at the graduate school of the University of Tokyo. (author)

  12. Recubrimientos depositados por cvd-fbr para protección a alta temperatura

    Jose Luddey Marulanda-Arevalo; Saul Castañeda-Quintana; Aduljay Remolina-Millan

    2013-01-01

    La deposición química de vapor por lecho fl uidizado (CVD-FBR) es una variante de la técnica de deposición química de vapor; que combina las ventajas de la activación térmica por calentamiento y el lecho fl uidizado. Los recubrimientos mediante CVD-FBR son ampliamente investigados y usados debido a la necesidad de proteger superfi cialmente los componentes que operan a altas temperaturas, el cual ha aumentado perceptiblemente. Además, tiene la ventaja de ser una técnica de bajo costo, puede c...

  13. Trends in theoretical calculation of dosimetry and gas production cross sections for FBR's, LWR's, and MFE's

    The energy ranges of prime interest for LWR's, FBR's, and MFE's are quite different, the low eV region for LWR's, the keV region for FBR's, and around 14 MeV for MFE's. Yet although the energy range is wide, the statistical model of nuclear reactions works well for most of the dosimetry and gas production needs. Even at 15 MeV where the predictions of the model are compared against experiments performed by Atomics International and Lawrence Livermore Laboratory the model works well

  14. Melting temperature of MOX fuel for FBR applications: TRANSURANUS modelling and experimental findings

    Calabrese, R., E-mail: rolando.calabrese@enea.it [ENEA, Reactor and Fuel Cycle Safety and Security Methods Section, via Martiri di Monte Sole 4, I-40129 Bologna (Italy); Manara, D., E-mail: Dario.MANARA@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Schubert, A., E-mail: Arndt.SCHUBERT@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Laar, J. van de, E-mail: Jacques.VAN-DE-LAAR@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Van Uffelen, P., E-mail: Paul.VAN-UFFELEN@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • An assessment of the solidus temperature models for MOX is presented. • The predictions of TRANSURANUS proved to be accurate in the investigated domains. • The model of Konno is of interest as it accounts properly for burn-up and MA effects. • Recent measurements support the need for a revision of the existing models. • The modelling of the O/M ratio effect needs further investigation. - Abstract: The paper is focused on the modelling of the melting temperature of mixed oxide (MOX) fuel for fast breeder reactors (FBRs). After a review of the models available in the TRANSURANUS (TU) code and in the open literature, their predictions were compared to an experimental dataset compiled from published measurements. The recommended model of TRANSURANUS was confirmed to be in good agreement with experimental data. A critical discussion of the comparison provided additional useful indications for the future development of the code and for the recommendations to the users involved in the analysis of the performance of fast reactor fuel. A special attention was given to the presence of minor actinides (MA), a topic of great importance for closure of the nuclear fuel cycle. In this frame, the code could be extended with the model of Konno in order to account for the presence of minor actinides. Finally, the review of the experimental data indicated the need for a reassessment of the effect of the oxygen-to-metal (O/M) ratio on the melting temperature in the low plutonium content domain of relevance to FBR fuel.

  15. The development of general assessment system for FBR cycle for practical use

    This research aims to develop a system in which aspects necessary for FBR cycle and overall comparison of evaluation items (economy, safety etc.) are evaluated quantitatively and objectively as a part of Nuclear Cycle development's research project of the FBR cycle for practical use. There are various methods in the decision-making support. In this particular situation, features of each method were evaluated based on the analysis of cases with each method. Subsequently we constructed overall evaluation method by combining Analytic Hierarchy Process (AHP), Multi-attribution Utility Function Method (MUF) and Cut-off Method. This method has variation in evaluation items, transparency in evaluation process and uncompensation. The six aspects of evaluation are economy, effectiveness of resource use, proliferation resistance, environmental effectiveness, safety, and research and development. The evaluation items and the evaluation index of each aspect were hierarchized and the evaluation structure was constructed. In the present study effect function for each evaluation index and pair comparison for examining significance of each item were utilized to select prospective systems for FBR cycle experimentally. The result confirmed reliability of our general assessment system as a decision-making support system for FBR system. (author)

  16. Design and construction of prototype FBR 'Monju' equipments by Fuji Electric

    The prototype FBR 'Monju' which has been constructed by Power Reactor and Nuclear Fuel Development corporation has attained the initial criticality on April 5, 1994. Fuji Electric was in charge of the fuel handling and storage facilities, radioactive waste treatment facilities and radiation monitoring facilities from their design, manufacture to installation in the prototype FBR 'Monju'. As for these equipments, the specifications and structures were decided through the results of the construction and operation of the experimental FBR 'Joyo' and its design and various development tests. The manufacture of the equipments was carried out in Kawasaki Works, the measuring and control system in Tokyo Works, and the power source board in Kobe Works. The installation works, the functional tests of the single equipments, and the integrated functional test of the whole prototype FBR 'Monju' were carried out, and the facilities were delivered in December, 1992. The outline, the basic specifications, the design and development of the main equipments, the manufacture of the equipments and so on of respective facilities are described. The installation works and the tests at the site are reported. (K.I.)

  17. Controlled synthesis and optical properties of BaFBr:Eu2+ crystals via ethanol/water solutions

    Graphical abstract: A facile and cost-effective approach for the controlled synthesis of BaFBr:Eu2+ crystals is introduced. The structures and morphologies of the obtained products are affected by the amount of water and ethanol in the solvent mixtures. Highlights: ► Precipitation route for preparing BaFBr nano and micro crystals in water/ethanol solvent mixtures. ► Controlled growth of BaFBr nano crystals by tuning the volume ratio of Ethanol/water. ► Luminescence properties after annealing at 200 °C are investigated. ► Short lifetimes of photoluminescence and photostimulated luminescence in BaFBr:Eu2+ nano crystals are presented. ► Shortened lifetimes in BaFBr:Eu2+ nano crystals demonstrate that they are promising materials for use in X-ray imaging systems. -- Abstract: BaFBr:Eu2+ crystals with different structures were successfully fabricated via a simple precipitation method using ethanol/water mixtures as solvents. The amount of ethanol in the solvent mixtures played a significant role in the formation of final products, enabling the well-controlled growth of the BaFBr crystals. A possible formation mechanism was proposed based on the results of controlled experiments. The phases and morphologies of the resulting samples were systematically investigated by X-ray diffraction (XRD), field-emission scanning electron microscopy (FESEM), transmission electron microscopy (TEM), high-resolution TEM (HRTEM), selected area electron diffraction (SAED) and elementary analysis. The optical properties of the annealed BaFBr:Eu2+ nano-cuboids were investigated using photoluminescence (PL), photo-stimulated luminescence spectroscopy (PSL) and kinetic decays. Faster decay behaviors demonstrate that these BaFBr:Eu2+ phosphors are promising materials for applications in optical storage fields. Furthermore, it is envisaged that this environmentally benign method can be extended to prepare other fluoride halides.

  18. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  19. Some difference of concepts between design guideline for FBR base isolation system and aseismic design guideline of LWR in Japan

    This paper deals with the concept and the relation of 'the Base Isolation System and FBR' to the Safety Criteria and the Guideline of the Aseismic Design of LWR in Japan. The Central Research Institute of Electric Power Industries have been working for FBR last several years. The author has been contribute to their works, and this is one of the subjects. He described his own idea obtained through the cooperative work with CRIEPI. (author)

  20. Development of computer code SAFFRON for evaluating breached pin performance in FBR's

    Ukai, Shigeharu; Shikakura, Sakae (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center); Sano, Yuji; Takita, Masami

    1994-07-01

    In order to evaluate the breached pin behavior in FBR, the breached pin performance analysis code SAFFRON was developed. Based on the results of run-beyond-cladding-breach test in FBR-II as a collaborative program between PNC and U.S.DOE, the following behaviors were taken into consideration; fuel sodium reaction product (FSRP) formation, resultant fuel expansion, breach extension of cladding and release of delayed neutron precursors into the coolant. Using 3-dimensional elastic analyses by finite element method, breached pin diameter increase is adequately predicted with the reduced Young's modulus of the breached fuel. The delayed neutron signal response in on-line diagnosis was evaluated in relation to the growth of FSRP and breached area enlargement. (author).

  1. Recubrimientos depositados por CVD-FBR para protección a alta temperatura

    Jose Luddey Marulanda-Arevalo

    2013-01-01

    Full Text Available La deposición química de vapor por lecho fluidizado (CVD-FBR es una variante de la técnica de deposición química de vapor; que combina las ventajas de la activación térmica por calentamiento y el lecho fluidizado. Los recubrimientos mediante CVD-FBR son ampliamente investigados y usados debido a la necesidad de proteger superficialmente los componentes que operan a altas temperaturas, el cual ha aumentado perceptiblemente. Además, tiene la ventaja de ser una técnica de bajo costo, puede controlar con relativa facilidad la composición del material depositado, permitiendo realizar recubrimientos con una orientación preferente que permite la obtención de intercaras con propiedades anisotrópicas; estos son depositados a bajas temperaturas y a la presión atmosférica.

  2. Experience in the use of FBR core component structural design criteria as applied to FFTF

    Hecht, S.L.

    1979-01-01

    User gained experience resulting from trial applications of proposed structural design guidelines for Fast Breeder Reactor (FBR) core components is presented. This work was done supporting the design analyses process for consumable core components for the Fast Test Reactor (FTR) of the Fast Flux Test Facility (FFTF). The proposed guidelines were found to be more comprehensive and generally easier to apply than those methods previously used. Component evaluation required a minimum amount of detailed inelastic analysis, primarily through the use of simplified inelastic analysis methods, as given in the guidelines. A major shortcoming of this draft criteria/guidelines is a lack of supporting irradiated material properties. Some areas of guidance given seems ambiguous and may be non-conservative, particularly those related to stress classification unique to FBR environments. Further verification of these areas appears to be in order.

  3. Application of the PSA method to decay heat removal systems in a large scale FBR design

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10-7/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  4. A study on the design method of the laminated rubber bearing for FBR

    The Central Research Institute of Electric Power Industry (Japan) has been carrying out the Demonstration Test and Research Program of the Seismic Isolation System for the Fast Breeder Reactor (FBR) under contract with the Ministry of International Trade and Industry (FY1987--FY1993). In this research program, development of the seismic isolation element, reliability evaluation of the seismic isolation system, and examination of a design-based earthquake have been conducted in order to develop and propose design guidelines of seismic isolation system for FBR plant. The purpose of this paper is to describe the test results concerned with structural design of the laminated rubber bearing, and to explain the construction of the design method for the laminated rubber bearing, which is one of the most important factors in the design guidelines of the seismic isolation system

  5. Myristylation alters DNA-binding activity and transactivation of FBR (gag-fos) protein.

    Kamata, N; Jotte, R M; Holt, J. T.

    1991-01-01

    FBR murine sarcoma virus (gag-fos) protein, a virally transduced Fos protein, exhibits decreased gene transactivation in comparison with the cellular Fos protein. Biochemical analysis suggests that myristylation of the virally encoded N-terminal gag region results in decreased DNA binding and transcriptional activation without affecting heterodimerization with Jun protein. These findings demonstrate that protein myristylation can modulate gene regulation by a DNA-binding protein.

  6. Development of inspection and repairing technology for FBR heat exchanger pipes

    A hybrid optical fiber scope and a laser processing head were combined with an eddy current testing unit for the inspection and repairing for Fast Breeder Reactor's heat exchanger pipes. The project proposes an effective cost reduction method for maintenance of FBR by extension of the heat exchanger's lifetime. By the end of March 2010, a prototype probing system for heat exchanger pipes will be completed. (author)

  7. Positron annihilation investigation of BaSrFBr:Eu by X-ray irradiation

    Lee, C. Y.

    2014-12-01

    The mechanical property of the BaSrFBr:Eu phosphor layer of X-ray image plates was investigated by using resolution (LP/mm) and coincidence Doppler broadening (CDB) positron annihilation as well as positron annihilation lifetime (PAL). The image plate samples of BaSrFBr:Eu phosphors in this experiment were irradiated by using hospital X-rays. The LP/mm values of the irradiated BaSrFBr:Eu image plates varied from 3.35 to 1.25 for up to 20,000 exposures. CDB positron annihilation and lifetime spectroscopy were used to analyze defect structures in the phosphor layer. Even when the LP/mm values were greatly decreased due to exposures, the S parameter and the lifetime ( τ 1, τ 2) values were almost constant with increasing number of exposures. A positive relationship existed between the SEM images and positron annihilation spectroscopy (PAS). According to the SEM images and the positron annihilation spectroscopy (PAS) results, measurements of the defects with PAS indicate that the image-plate phosphor can be safely used for hospital X-rays in the course of diagnostic radiography at an average rate of 20,000 times for one year.

  8. Status of MOX fuel development for the FBR system in JAEA

    From the damage by the Great Eastern Japan Earthquake on March 11, 2011, the Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) has been restored. The mission of the NCL is to establish the FBR fuel cycle based on the MOX fuel. The NCL has developed and demonstrated the optimization of the interface technology between the reprocessing and the MOX fabrication. The MOX fuel fabrication technology has been developed in the Plutonium Fuel Development Center (PFDC) of the NCL. Considering smooth transition from the LWR era to the FBR era, it is essential to use degraded Pu recovered from high burn-up LWR fuels or LWR-MOX. For the FBR MOX fuel fabrication, an original procedure named the Short Process has been designed and the test has been conducted since 2008. Through tests, 81 fuel assemblies for the Monju have been fabricated. Pu has been handled in the NCL under the 3S-philosophy, they are Safeguards, Security and Safety. For the future fuel, the fabrication technology development on the annular pellet as well as the study on the fundamental physical property has been carried out. (author)

  9. Defect analysis of BaSrFBr:Eu irradiated by X-ray

    The mechanical property of the BaSrFBr:Eu phosphor layer of X-ray image plates was investigated by using image quality (IQ), resolution (LP/mm), and coincidence Doppler broadening (CDB) positron annihilation. The screen samples of BaSrFBr:Eu phosphors were irradiated with hospital X-rays in the course of diagnostic radiography at an average rate of 20,000 times per year and were used for various periods of time. The LP/mm values of the irradiated BaSrFBr:Eu image plates varied between 2.4 and 2.0 for three years while the IQ values varied between 35 and 11 over the same period. CDB positron annihilation spectroscopy was used to analyze the defect structures in the phosphor layer. The S parameter values increased in correlation with increased exposure time, which indicated that more defects were generated. There was a positive relationship between the IQ and S parameters. Measurements of the defects indicate that most of the defects were likely to have been generated by the X-ray radiation.

  10. Methodology for seismic analysis of FBR core assembly using variable added damping

    It is necessary to consider the fluid-solid coupling effect due to the interaction between coolant and core assembly when analyzing the seismic performance of the FBR core assembly. The added damping was treated mostly as a constant in previous researches. In fact, the effect on assemblies from the coolant depends strongly on the gap between the core assemblies, and the damping should be considered as a variable. In order to simulate the vibration of the core assembly more accurately, the methodology for the seismic analysis of FBR core assemblies using variable added damping was studied. In this paper, the seismic analysis model of one single row of core assemblies (5 assemblies) of FBR was established. By comparing the two kinds of added damping models, the constant and variable ones respectively, the results show that the seismic analysis of the core assemblies with variable added damping is feasible and effective. Meanwhile, the simulation method used in this paper can obtain more precise approximation of the vibration of the core assembly and lays the foundation for more realistically simulating the seismic response of reactor core assemblies. It also helps to reduce the conservative margin of the structural design and is meaningful in engineering application. (authors)

  11. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D20- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

  12. Seismic analysis of nuclear island connected buildings for 500 MWe PFBR (Part-1 Mathematical modelling and free vibration characteristics)

    PFBR, prototype fast breeder reactor is a 500 MW(e) fast reactor meant for power generation on commercial scale. It is the first breeder reactor in India. Various safety related buildings in the complex of PFBR are interconnected to enhance safety, strength and economy, resulting in large shear wall framed structure, referred as Nuclear Island Connected Buildings (NICB). NICB houses reactor assembly, primary and secondary sodium circuits, steam generators, spent fuel storage pool and many other safety related systems. It is embedded to a depth of 20.6m below finished ground level. For analysis, a three-dimensional finite element model of the structure consisting of plate elements, beam elements, mass elements and springs along with dampers is developed. Main Equipment are idealised as stick models and liquid in storage pool too is idealised as per ASCE: 4-98. Effect of soil structure interaction is accounted in the dynamic analysis by impedance method of ASCE: 4-98. The modeling aspects of the analysis and results of free vibration analyses with varying foundation modulus during dynamic condition are presented in this paper. (author)

  13. Profiles of facilities used for FBR research and testing

    This document contains a concise up-to date of supporting ''Liquid Metal Fast Breeder'' facilities submitted by countries (Federal Republic of Germany, Italy, France, Belgium, Netherlands, UK, USA, USSR and Japan) and international organizations. It has the purpose of providing an over-view of the facilities currently used or under construction and of the type of experiments that can be conducted therein

  14. Fast breeder reactor core concept consistent with fuel cycle system during the transition period from LWR to FBR cycles in Japan

    During the transition period from the LWR cycle to a FBR cycle, the Pu needed for FBR start-up will be obtained from the next reprocessing facility. Therefore, from the economic viewpoint, it would be better if the FBR core in the next reprocessing facility had a lower Pu inventory. A design concept study was carried out for an FBR core with a lower Pu inventory that is consistent with the fuel cycle system during the transition period. The Pu from LWR spent fuel will be used for FBR start-up at least for the initial core cycle and succeeding first few cycles during the transition period. An FBR core loaded with Pu from LWR spent fuel has higher burnup reactivity than one loaded with Pu from FBR multi-recycling fuel composition. The increased burnup reactivity may reduce the cycle length of the FBR core. We chose loading of minor actinides (MAs) into the FBR MOX fuel as the countermeasure to the increased burnup reactivity from the viewpoint of utilizing nuclear characteristics of the MAs. The maximum MA content in the MOX fuel was set to 5% on the basis of irradiation test results. MAs recovered from the LWR spent fuel that provides Pu for FBR start-up are loaded into the initial loading fuel assemblies and exchanged fuel assemblies for several cycles until equilibriumis reached. The average MA content of the initially loaded fuel was assumed to be 3%, and that of the exchanged fuel was assumed to be 5%. The core performance including burnup characteristics and reactivity coefficient were also evaluated, and it was confirmed that the transient core from the initial loading until the equilibrium cycle for loaded Pu from LWR spent fuels could maintain performance neary equal to that of an FBR multi-recycling core. (author)

  15. BaFBr:Eu2+ nanophosphor-SiO2 hybrid entrapped in Anodise Alumina membrane pores array

    Sol–gel template method has been used to prepare BaFBr:Eu2+ nanophosphor-SiO2 hybrid entrapped within the nanopores array (of about 200 nm size) of a comercial anodized alumina (AA) membrane. Structural and morphological measurements using electron microscopy (SEM) and X-ray diffraction (XRD) have shown the presence of the BaFBr:Eu2+ nanophosphor in the silica xerogel entrapped within the nanopores array; photoluminescence and X-ray excited luminescence measurements have shown Eu2+ luminescence at 395 nm accompanied by a broad band due to AA membrane. The method assures a relatively uniform spreading of the BaFBr nanophosphor into the AA membrane pores array without the nanoparticles agglomeration. Preliminary imaging tests have shown a spatial resolution in the micrometer range and even in the submicrometer range can be expected. As BaFBr:Eu2+ is a very efficient X-ray phosphor the material might be used as X-ray micro-imaging detector. - Highlights: • Sol–gel method was used to prepare Eu-doped BaFBr nanophosphor embedded in SiO2 matrix. • Anodized alumina membrane nanopores array were filled by the nanophosphor-SiO2 hybrid. • Photo and X-ray luminescence spectra showed Eu2+ ions luminescence at 395 nm. • Preliminary imaging tests have shown a spatial resolution in the micrometer range

  16. Proposal of a nuclear cycle research and development plan in Tokai works. The roadmap from LWR cycle to FBR cycle

    The Generation-II Project Task Force Team has investigated a research and development plan of a future nuclear fuel cycle in Tokai works for about three months from December 19, 2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R and D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. Establishment and advancement of 'the tough nuclear fuel cycle'. Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste. And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology. (author)

  17. Present condition of survey research on actualization strategy of fast breeding reactor (FBR) cycling. General outlines on the research

    The Japan Nuclear Cycle Development Institute (JNC) started the survey research on actualization strategy of FBR cycling under cooperation of related organizations such as electric business company and so on, on July, 1999. The research aims at preparation of technical system to establish the FBR cycling for a future main energy supply source by extracting an actualization picture maximum activated advantages originally haven by the FBR cycling and by proposing a developmental strategy flexibly responsible to diverse needs in future society. Here was reported on effort state of its phase 1 (two years between 1999 and 2000 fiscal years). In the phase 1, it was planned to perform research and development shown as follows: 1) Extraction of actualization candidate concept on the FBR cycling under a premise of safety security and a viewpoint of evaluation on economics, resource effective usage, environmental loading reduction, and nuclear dispersion resistance by conducting investigation and evaluation of wide technical choices adopting innovative techniques, and 2) Embodiment of a research and development program of phase 2 (from 2001 to 2005 fiscal years) by investigating some technical subjects important for selection of research and development program aiming at actualization and its candidate concept on the FBR cycling. (G.K.)

  18. Development of knowledge-based operator support system for steam generator water leak events in FBR plants

    A knowledge engineering approach to operation support system would be useful in maintaining safe and steady operation in nuclear plants. This paper describes a knowledge-based operation support system which assists the operators during steam generator water leak events in FBR plants. We have developed a real-time expert system. The expert system adopts hierarchical knowledge representation corresponding to the 'plant abnormality model'. A technique of signal validation which uses knowledge of symptom propagation are applied to diagnosis. In order to verify the knowledge base concerning steam generator water leak events in FBR plants, a simulator is linked to the expert system. It is revealed that diagnosis based on 'plant abnormality model' and signal validation using knowledge of symptom propagation could work successfully. Also, it is suggested that the expert system could be useful in supporting FBR plants operations. (author)

  19. Suitability of a thermal design method for FBR oxide fuel rods

    To study suitability of the thermal design method for fast breeder reactor (FBR) oxide fuel rods, the B14 irradiation test with four fuel rods was carried out in the experimental fast reactor 'JOYO'. Pellet-cladding gap width and O/M ratio of oxide fuels were specified as experimental parameters. In addition, by taking into account the actual design conditions for FBR oxide fuel, the conditions in the B14 irradiation test, i.e. linear power and cladding temperature, were planned to include the hottest design conditions. The maximum Pu content and the maximum Am content of fuel pellet were 31 wt% and 2.4 wt%, respectively. The B14 fuel rods were irradiated with the maximum linear power of ∼ 47 kW/m in the test. After irradiation, ceramography samples were taken from the axial position of each fuel rod where the fuel centerline temperature reached the maximum during irradiation. The result was that the influences of both pellet-cladding gap width and O/M ratio on the fuel restructuring were observed, but the fuel melting was not observed. In addition, thermal analysis code 'DIRAD' would be suitable to evaluate the thermal behavior of oxide fuels containing several percent Am, from the result of verification by using the result of the B14 irradiation test. Moreover, from the computation result of DIRAD, the power to melt for the B14 oxide fuels was evaluated as 55-57 kW/m. It could be mentioned that the margin of the conventional oxide fuel design would be at least 10 kW/m at the transient. Consequently, the margin to the criterion in the thermal design would be suitable and the fuel melting would be prevented under the conditions designated in the conventional FBR design.(author)

  20. Storage and disposal of high-level radioactive waste from advanced FBR fuel cycle

    Waste management of fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P and T) technology was investigated by focusing on thermal constraints due to heat deposition from waste in storage and disposal facilities including economics aspects of those facilities. Partitioning of minor actinides (MAs) and heat-generating fission products in high-level waste can enlarge the containment ratio of waste elements in the glass waste forms and shorten predisposal storage period. Though MAs can be transmuted in FBRs or dedicated transmuters, heat-generating fission products are difficult to be transmuted; they are partitioned and stored for a long time before disposal. The disposal concepts for heat-generating fission products and remainders such as rare-earth elements depend on storage period that ranges from several years to several hundreds of years. Short-term storage results in small size of storage facilities and large size of repositories, and vice versa for long-term storage. This trade-off relation was analyzed by estimating repository size as a function of storage period. The result shows that transmutation of MAs is essentially effective to reduce repository size regardless to storage period, and a combination of P and T can provide a smaller repository than the conventional one by two orders of magnitude. The cost analysis for waste management was also made based on rough assumptions on storage, transportation and repository excluding cost for introducing P and T that are still under evaluation. Cost of waste management for FBR without P and T is 0.25 Yen/kWh that is slightly smaller than that for LWR without P and T, 0.30 Yen/kWh. The introduction of MA transmutation to the FBR results in cost of 0.20 Yen/kWh, and full introduction of P and T provides the smallest cost of 0.08 Yen/kWh. (author)

  1. Structural integrity evaluation method for overheating rupture of FBR steam generator tube

    This paper describes a structural integrity evaluation method for a SG tube of FBR in case of sodium-water reaction and creep rupture tests to obtain the strength of the tube material. In the SG of FBR, if intermediate size of water/steam leak (1-2 kg s-1) would occur from a tube, it could cause overheating rupture of the multiple tubes surrounding the initially failed tube due to generated sodium-water reaction heat. In the ultra-high temperature condition, the creep strength of the material is one of the dominant factors for failure behavior. Accordingly, we tried to apply the creep failure criterion for the overheating rupture of the SG tube. The creep rupture tests have been performed at ultra-high temperature conditions ranging from 1223.2 to 1323.2 K. The test material is 'Mod.9Cr-1Mo steel' which is one of the candidate materials for the tubes of the future SG of FBR. The test results have shown that tube rupture depends on the creep strength of the material; hence, instantaneous rupture does not occur even if the stress exceeds the design value of ultimate tensile strength. The test data have been suitably expressed using the Larson-Miller Parameter, and a structural integrity evaluation method based on the sum of the use-fraction associated with the creep damage has been proposed. Based on this method, the structural integrity of the tube in the sodium-water reaction flame has been evaluated. The results show that it is important to detect the initial leak of the tube within a short period and to reduce the steam pressure more rapidly by SG blowdown

  2. Consistency evaluation of JUPITER experiment and analysis for large FBR cores

    A series of critical experiments for study of large FBR cores, JUPITER, was analyzed with the latest analytical methods. These results were evaluated from various physical viewpoints by means of comparison with other cores or other nuclear characteristics by full use of sensitivity analysis, effect of different nuclear data libraries and application of most-detailed analytical tools. It is concluded that the JUPITER experiments and analytical results possess sufficient consistency on the whole, though there is some room for further improvements. The proper use of JUPITER data will enhance the accuracy and reliability of design work for the large FBRs. (author)

  3. Temperature monitoring and leak detection in sodium circuits of FBR using Raman distributed fiber optic sensor

    This paper discusses the fiber optic temperature sensor based leak detection in the coolant circuits of fast breeder reactor. These sensors measure the temperature based on spontaneous Raman scattering principle and is not influenced by the electromagnetic interference. Various experiments were conducted to evaluate the performance of the fiber optic sensor based leak detection using Raman distributed Temperature Sensor (RDTS). This paper also deals with the details of fiber optic sensor type leak detector layout for the coolant circuit of FBR, performance requirement of leak detection system, description of the test facility, experimental procedure and test results of various experiments conducted. (author)

  4. Coprocessing solvent-extraction flowsheet studies for LWR and FBR fuels

    Coprocessing solvent extraction studies using irradiated LWR and FBR fuels have indicated the need for an efficient feed clarification. A potentially useful filtration method for fulfilling this need has been demonstrated. Conditions necessary for the satisfactory use of unstabilized hydroxylamine nitrate (HAN) or nitrous acid as the reducing agent for Pu(IV) during reductive stripping operations have been defined. Both partial partitioning and total costripping operations have been demonstrated. In addition, solvent degradation product measurements have been made, and the effect of the presence of DBP during uranium-plutonium stripping operations has been determined

  5. Comparative analysis of coolants for FBR of future nuclear power

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  6. Investigation of 144Pr measurements for determination of residual Pu in FBR leached hulls

    The measurement of plutonium in leached hulls arising from FBR fuel reprocessing is required for plant control, accountancy and safeguards. At DNPDE these measurements are carried out on the batch of hulls prior to bulking into 200 l drums for retrievable storage and ultimately further treatment for plutonium recovery. The experience to date has related to the use of neutron interrogation using sealed tube neutron generators as the irradiation source. The supply of replacement sealed tubes has become difficult. It was therefore decided to consider, amongst other techniques, the measurement of 144 Pr gamma emission as a possible alternative. The technique has been extensively used for thermal reactor fuels on a routine basis. There has also been a limited amount of work reported from Cap La Hague on applying the technique to FBR fuels as part of an experimental programme. This paper therefore describes the work done at DNPDE which evaluated the technique for use on a batch of hulls arising from a PFR fuel reprocessing campaign. (author)

  7. Biotemplating of BaFBr:Eu{sup 2+} for X-ray storage phosphor applications

    Kostova, M.H. [Department of Materials Science and Engineering - Glass and Ceramics, University of Erlangen - Nuremberg, Martensstr. 5, D-91058 Erlangen (Germany); Batentschuk, M. [Department of Materials Science and Engineering - Materials for Electronics and Energy Technology, University of Erlangen - Nuremberg, D-91058 Erlangen (Germany); Goetz-Neunhoeffer, F. [GeoZentrum Nordbayern, Mineralogy, University of Erlangen - Nuremberg, D-91058 Erlangen (Germany); Gruber, S. [Department of Materials Science and Engineering - Glass and Ceramics, University of Erlangen - Nuremberg, Martensstr. 5, D-91058 Erlangen (Germany); Winnacker, A. [Department of Materials Science and Engineering - Materials for Electronics and Energy Technology, University of Erlangen - Nuremberg, D-91058 Erlangen (Germany); Greil, P. [Department of Materials Science and Engineering - Glass and Ceramics, University of Erlangen - Nuremberg, Martensstr. 5, D-91058 Erlangen (Germany); Zollfrank, C., E-mail: cordt.zollfrank@ww.uni-erlangen.de [Department of Materials Science and Engineering - Glass and Ceramics, University of Erlangen - Nuremberg, Martensstr. 5, D-91058 Erlangen (Germany)

    2010-09-01

    The design of hierarchically patterned novel structures by replicating the cellular tissue of wood has recently attained increasing interest. X-ray storage phosphor BaFBr:Eu{sup 2+} is manufactured via vacuum assisted repeated infiltration of wood tissue (Pinus sylvestris). A submicrometer precipitate is formed via wet chemical reaction of NH{sub 4}F, BaBr{sub 2}.2H{sub 2}O and EuCl{sub 3}.6H{sub 2}O in methanol. According to scanning electron microscopy (SEM) and energy dispersive X-ray analysis (EDX), the original wood cell walls are filled with the precipitate and completely transformed into BaFBr struts after sintering at 800 deg. C. The optical properties of the biomorphous phosphor microstructure are determined by photoluminescence spectroscopy (PL) at room temperature, photo-stimulated luminescence spectroscopy (PSL) and cathodoluminescence spectroscopy (CL) in the SEM. A broadening of the PSL peak is observed and ascribed to the incorporation of calcium impurities present in the pine wood tissue. The potential of biotemplates for generating highly oriented and optically isolated {mu}m- and sub-{mu}m matrix of X-ray storage phosphor material is illustrated.

  8. A study on the reprocessing of spent FBR-fuel by ion exchange. 2

    In order to develop an economically efficient wet separation process other than solvent extraction for reprocessing spent FBR-fuel (MOX fuel), we have investigated the possibility of an advanced ion exchange process. Based on the results of fundamental research and the fruits of this research in last year, the proposed FBR-fuel reprocessing process which consists of anion exchange separation and extraction chromatography separation has been studied quantitatively from the engineering aspect. The plant concept, construction cost, applicability of this process were investigated and preliminarily evaluated. The proposed process was improved to reduce the amounts of operation solution and waste generation, and to enhance the properties of the impregnation adsorbents for MA separation. The mass balance including waste generation in main processes was evaluated. The operation flow sheets for each process were drawn. The main machines were conceptually designed. Furthermore, conceptual design for the reprocessing plant using ion exchange and extraction chromatography was executed and the installation layouts of the machines, equipment and facilities were examined and designed. Based on the research results, the construction cost for the reprocessing plant was estimate and compared with the existing PUREX plant. Finally, the subjects resulted from the introduction of the ion exchange process were extracted and the considerations for solving these subjects were also indicated. (author)

  9. Photostimulated luminescence properties of BaFBr:Eu under ion irradiation

    The photostimulated luminescence (PSL) properties of the phosphor BaFBr:Eu after ion beam irradiation was analyzed; in particular, the PSL intensity dependent on ion fluence. The PSL intensity increased linearly with the ion fluence up to 1012 ions/cm2, and subsequently decreased gradually. The ion fluence dependence was observed to be similar among samples containing different F centers or different Eu concentrations. The fluence dependence was quantitatively analyzed based on a trapping model, in which competition between the trapping processes to storage centers and radiation defects is assumed; the model explained the experimental data quantitatively. The results indicate that radiation defects influence the PSL properties via the trapping of photostimulated electrons. - Highlights: • Photostimulated luminescence (PSL) behavior of BaFBr:Eu2+ is analyzed for ion irradiations. • PSL intensity as a function of fluence is described based on a trapping model. • It is shown that the decrease in the PSL intensity is due to the trapping of photostimulated electrons to irradiation defects. • It is shown that the trapping process is quantitatively different for the sample with different Eu2+ concentrations. • It is strongly suggested that the range of the photostimulated electrons is larger than 10 nm

  10. Development of cost-benefit analysis system for FBR research and development

    In this research, we performed investigation and evaluation on cost-benefit, and made system construction of basic concept and conceptual design, prior to the system construction which performs (1) evaluation of cost-benefit of FBR development within the country, and (2) evaluation of cost-benefit for every item for R and D relating to the FBR development. As for the methods for cost-benefit, it shall be suitable to assume that success rate differs depending on the future scenario such as fuel prices and to employ discount rates, considering the point of long-term R and D investment. We also pointed out that establishment of discount rates is the biggest issue which influences the results of evaluation of cost-benefit then. When we grasp the effects, it is required to include the effects such as contribution to environmental benefits and energy security, or improvement of safety, other than real generation merits, as it is a public investment. Therefore we made investigations and developments on the way to understand the effects qualitatively. Based on the results, we made a proposal on application to the system, and then showed evaluation flows on the cost-benefit system. We also pointed out that it is a future task to be considered to establish fuel cost scenario or input data or parameters such as value of statistical life or CO2 reduction, as they are primary factors which influence evaluation results or system reliabilities. (author)

  11. Distinctive features of proposed technical guidelines for the design of seismically isolated fast breeder (FBR) plants

    The application of seismic isolation technology to fast breeder reactor (FBR) plants is expected to reduce earthquake load to both the building and apparatus of the plants. It is also expected to facilitate the development of a rational approach to all phases of the earthquake-proof design work. Seismic isolation technology has already been applied painstakingly to non-nuclear industrial facilities and civil structures. The design method has been partially verified for the specific applications. However, the application of the technology to nuclear power reactor plants requires greater reliability than needed for ordinary buildings. Under request from the Ministry of International Trade and Industry (MITI) of Japan, the Central Research Institute of the Electric Power Industry (of Japan) has performed verification tests on seismic isolation technology, and worked toward establishing and proposing technical guidelines for FBR plant design. This project has been performed over seven years, from 1987 to 1993. Results of previous studies and data of the verification tests conducted in this project are reflected in the proposed guidelines presented here. Major features of the proposed guidelines are outlined below

  12. Development of a new tube-to-tubesheet welding type for FBR's heat exchangers

    The intermediate heat exchanger for exchanging heat between primary and secondary sodium and the steam generator are required to guarantee the performance and the reliability of construction over long term under the environment of high temperature sodium peculiar to a FBR. With the increase of power output of FBR plants, the number of heating tubes and the size of tube plates of the intermediate heat exchangers of shell and tube type increase. In order to improve the reliability and the production process of the tube to tube plate welding, a new method was developed, according to which heating tubes are inserted into tube plate holes by about their thickness, and the welding of perfect penetration is carried out internally with an automatic TIG welder. At the same time, in order to confirm the reliability of the welded joints by this method, the tests for evaluating the strength for short time and long term were carried out. It was confirmed that the satisfactory performance of the welded joints was able to be obtained, and the method would be applicable to actual heat exchangers. As for the nondestructive inspection of welded joints, the radiographic method was established, and ultrasonic and eddy current flaw detection methods are being developed now. (Kako, I.)

  13. Study on decay heat removal characteristics by natural circulation for top entry loop-type FBR

    The thermal-hydraulic circulation characteristics of a top entry loop-type FBR are studied experimentally with a 1/8 scaled three-loop model using water as the working fluid. The model is designed to simulate reactor decay heat removal operations including natural circulation mode as well as three types of decay heat removal systems, a direct reactor auxiliary cooling system (DRACS) with a direct heat exchangers immersed in a hot plenum of a reactor vessel (Immersed type DRACS), a DRACS with DHXs penetrating from a hot plenum to a cold plenum of a reactor vessel (Penetrating type DRACS) and a primary reactor auxiliary cooling system (PRACS). The feasibility of reactor decay heat removal operations including natural circulation mode of a top entry loop-type FBR is confirmed based on the experimental results. The natural circulation performance of the Immersed type DRACS is compared with that of the PRACS for the water test results. The results are used to evaluate the accuracy of the one-dimensional flow-network analysis codes and the multi-dimensional thermal-hydraulic analysis code. The analytical results are generally in good agreement with the measurements. (author)

  14. Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel

    Creep fatigue crack growth contributes to the failure of FBR reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading by simplified JNC method. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions by FINAS code. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS, NDE and shallow cracks) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. Circumferential cracks become more hazardous than longitudinal cracks in the reactor. The total crack growth of the largest PTS crack is about 2mm after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life

  15. Distinctive features of proposed technical guidelines for the design of seismically isolated fast breeder (FBR) plants

    Ishida, Katsuhiko; Yabana, Shuichi [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Earthquake Engineering Group; Shibata, Heki [Yokohama National Univ., Kanagawa (Japan)

    1995-12-01

    The application of seismic isolation technology to fast breeder reactor (FBR) plants is expected to reduce earthquake load to both the building and apparatus of the plants. It is also expected to facilitate the development of a rational approach to all phases of the earthquake-proof design work. Seismic isolation technology has already been applied painstakingly to non-nuclear industrial facilities and civil structures. The design method has been partially verified for the specific applications. However, the application of the technology to nuclear power reactor plants requires greater reliability than needed for ordinary buildings. Under request from the Ministry of International Trade and Industry (MITI) of Japan, the Central Research Institute of the Electric Power Industry (of Japan) has performed verification tests on seismic isolation technology, and worked toward establishing and proposing technical guidelines for FBR plant design. This project has been performed over seven years, from 1987 to 1993. Results of previous studies and data of the verification tests conducted in this project are reflected in the proposed guidelines presented here. Major features of the proposed guidelines are outlined below.

  16. Production enhancement and quality degradation of Pu produced in FBR blankets

    Full text: For smooth deployment of FBR in near future, securing economy and non-proliferation are pivotal factors. This study has two main objectives related to these two key factors. One is to enhance Pu production efficiency in blanket region of fast breeder reactor core. The other is Pu composition degradation to improve the proliferation resistance of FBR fuel cycle. The former contributes to reduce amount of blanket fuels to be fabricated and reprocessed to gain the same quantity of Pu, consequently it is expected to improve the fuel cycle economy. The composition of Pu generated in FBR blanket generally has very high quality, namely, equal to weapon or super grade. The latter objective is to deteriorate the quality of generated Pu by adjusting neutron spectrum in the blanket region and the initial composition of blanket fuels. A conceptual core design of MOX fueled, sodium cooled FBR with 1500MWe rating is performed by using SLAROM and CIRATION code with nuclear data set prepared for fast reactors of JFS-3-J3.2. The initial fuels in the active core region contains 5% of minor actinides produced in UO2-fueled PWR with enrichment of 4.1% and burnup of 43GWd/t after 10 years cooling. A index FIR (fissile inventory ratio at EOC and BOC) is used to measure the net fissile balance that is suited for fuel cycle mass balance analysis in addition to the BR (breeding ratio) in the standard definition. The moderator material used to tailor the neutron spectrum in the blanket region is ZrH. The depleted uranium pins in the radial blanket are replaced by ZrH pins in the volume fraction range from 0% to 30%. By increasing ZrH pins, the Pu production per unit mass of UO2 in blanket increased from 0.08kg-Pu/kg-UO2 to more than 0.1kg-Pu/kg-UO2. The FIR also slightly improved by replacing small fraction of UO2 by ZrH pins however it turned for the worse in the higher range more than 5% of ZrH. Loading ZrH into blanket fuel assembly drastically affects the Pu composition at

  17. Physics aspects of metal fuelled fast reactors with thorium blanket

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core

  18. Human Development in Japan and Abroad Using the Prototype FBR Monju Towards the Next-Generation Age

    Aiming at contributing to human development in Japan and abroad towards the next-generation, the International Nuclear Information and Training Center (INITC) has been working on educational training which consists of various educational training programs using the Fast Reactor Training Facility (FRTF) including prototype FBR 'Monju'. (author)

  19. Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

    As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

  20. Study on the FBR cycle introduction scenario III. A new harmonized concept of hard energy path and soft energy path in the future hydrogen society

    This report provides the results of our investigation about the new harmonized concept of hard energy plan and soft energy path in the future hydrogen society, as a part of FBR cycle introduction scenario study in the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System'. First, we considered that an environmental preservation policy and hydrogen utilization might be the 21st century, then, investigated the supply-and-demand status of resources and energy in hydrogen society. Furthermore, the harmonized image of a nuclear fuel cycle and soft energy paths (distributed power supplies, fuel cell vehicle, chemical plant, etc.) was investigated. Nuclear fuel cycle is considered to contribute to the supply of the platinum group metals whose the increase in demand will be expected in hydrogen society in addition to production of the CO2 free energies (electric power, hydrogen, etc.) by nuclear energy. And the range of the supply-and-demand balance of the platinum group metals recovered from the reprocessed waste was calculated. Furthermore, its radioactivity characteristics and the problem on the utilization of the recovered elements was evaluated. In order to propose the supply system of the resource and energy preserving the environment. We are going to study the infrastructure for the supply of recovered elements and CO2 free energy. (author)

  1. Design and development of thick plate concept for rotatable plugs and technology development for future Indian FBR

    Highlights: ► Thick plate concept and design for rotatable plugs of future Indian FBR described. ► Salient aspects of 800 mm thick plate welding technology development presented. ► Non-destructive examination (NDE) and destructive testing aspects discussed. ► Ultrasonic wave is attenuated less in narrow gap carbon steel weld than base metal. ► Recommendation made regarding RCC-MR code rules for NDE of carbon steel weld. - Abstract: Prototype fast breeder reactor (PFBR) is at an advanced stage of construction in Kalpakkam, India. Top shield (consisting of roof slab and rotatable plugs) of PFBR is of box type construction. It is a Class-1 component being a part of the primary leak-tight boundary of the reactor assembly. In the future commercial fast breeder reactor (CFBR), it is planned to adopt thick plate concept for rotatable plugs. The thickness of the plates required for the rotatable plugs of CFBR was arrived at based on finite element analysis considering mechanical and thermal loads. The main advantage of using thick plate for the rotatable plugs is that it eliminates the possibility of lamellar tearing which exists in the box type construction. The successful realisation of this concept necessitates the indigenous development of thick plate narrow gap welding technology. Hence a technology development programme for realising 800 mm thick narrow gap welds using submerged arc welding process was undertaken. Three numbers of welded joints were made to demonstrate the successful development of thick plate narrow gap welding technology. Repair procedure for the weld was also established. Non-destructive examination and destructive testing were carried out and the results were analysed critically. The attenuation of ultrasonic wave in narrow gap weld as compared to that in the base metal was examined. This paper discusses the need and conceptual design of thick plate concept for rotatable plugs in CFBR, the approach adopted for technology development of

  2. Training report of the FBR cycle training facility for Monju Restarting

    The FBR Cycle Training Facility consists of the sodium handling training facility and the maintenance training facility has been used since September in 2000. At the sodium handling training facility, the trainees are able to study widely sodium handling technologies, which are inherent technologies of fast reactors, such as sodium fire extinguishing, chemical and physical properties of sodium, sodium loop operating skill, counter training for sodium piping leakage, etc.. On the other hand, maintenance training courses not only for 'MONJU' related inherent maintenance technologies but also for general maintenance skills are conducted at the maintenance training facility. So far, 77 eight-unit sodium training courses and 39 nine-unit maintenance training courses have been performed in preparation of Monju restarting and the total number of trainees is 888. (author)

  3. Studies of the effects of fuel EOS uncertainties on FBR disassembly energetics

    A principal source of uncertainty in the energetics of FBR core disassembly is the lack of mechanical and thermophysical data on fresh and irradiated fuel under the conditions of interest. The consequences of uncertainties are analysed in two areas: (i) the equation of state (EOS) or irradiated fuel and (ii) the specific heat of molten fuel. The current UK understanding of the role of fission products in the postulated disassembly phase of an HCDA is outlined giving particular emphasis to the possible effects of pre-disassembly heating. The authors draw as far as possible on the rather sparse experimental data and indicate where further work would be most useful. Further they discuss arguments suggesting there exists substantial uncertainty in the currently-accepted values of the specific heat of molten fuel, and show that this lack of knowledge implies that current estimates of accident excursion yield could be exaggerated by more than a factor of two. (author)

  4. Tritium measurement using a photo-stimulable phosphor BaFBr(I):Eu2+ plate

    Tritium measurement is indispensable for the fuel- processing systems of deuterium-tritium (DT)-fusion facilities. A new approach to detect tritium in regions deeper than the escape depth of beta-rays from tritium is being developed using an imaging plate (IP). The measurement principle of this approach is to observe bremsstrahlung X-rays induced by the tritium beta-rays. An IP made of europium-doped BaFBr(I), a photostimulated luminescence (PSL) material, is a two-dimensional radiation sensor. In the present study, the characteristics of this IP for measuring tritium by detecting bremsstrahlung X-rays, in particular a fading effect and the energy dependence of PSL sensitivities, are examined. (orig.)

  5. CFD analysis of coolant channel geometries for a tightly packed fuel rods assembly of Super FBR

    Guo, Rui, E-mail: guorui@asagi.waseda.jp; Oka, Yoshiaki

    2015-07-15

    Highlights: • Supercritical water heat transfer is validated against the experimental data. • Thermal hydraulic performance of three coolant channel cross-sectional geometries are analyzed. • Geometry B is superior to the other two geometries. - Abstract: This paper presents the CFD investigation on three cross-sectional geometries of coolant channels in the newly designed tightly packed fuel rods assembly for high breeding of a supercritical-pressure light water cooled fast breeding reactor (Super FBR). The calculations addressed the turbulence models and mesh conditions validated against the experimental data. The cladding temperatures and pressure drop are compared. It is concluded that geometry B (triangle with round corner) is superior to the other two (round and triangle with sharp corner) due to its excellent thermal hydraulic characteristics.

  6. Measurement of deformation of FBR fuel assembly wrapper tube by an innovative technique

    An innovative technique to measure the deformation of irradiated wrapper tube for FBR fuel assembly was developed, and installed in the hot cell of the Fuels Monitoring Facility in JAEA. In order to confirm the performance of this instrument, a number of measurements were carried out on the wrapper tubes irradiated to high fluence in the experimental fast reactor Joyo. In the instrument used until now, only three face to face distances in the hexagonal wrapper tube have been measured along the axial direction. On the other hand, in the instrument developed in this technique the face to face distances could be continuously measured along the lateral direction on the outer surfaces of wrapper tube. Using data obtained by this technique, the detailed analyses of deformation can be done throughout a whole wrapper tube. (author)

  7. Development of innovative system and technology on MOX fuel production for FBR

    New technologies and machines were innovatively developed for realizing high speed MOX fuel production for FBR. Outstandingly different point from former approaches is adding theoretical analysis of very basic boiling feature which gives scientific basis in the system design. The system to be improved consists from the granulation, pellet molding, sintering, and O/M control. It is effective for suppressing the overflow, which appears in the de-nitration process, to keep the D/H ratio larger than 3 for instance, where D is the inner diameter of vessel and H the liquid depth in the vessel. In the granulation experiments, averaged particle diameter 630 μm, and flowability 91 were the best result. The collecting rate of UO2 granules from UO2 powder reached 97 %. The die wall lubrication method brought us sure result to fabricate mechanically strong pellets by increasing the molding pressure. An attempt to assemble an efficient layout of these apparatus is ongoing. (author)

  8. Design of a new MOX powder transport packaging to support FBR cycle development mission

    The Japan Nuclear Cycle Development Institute (JNC), which plays a leading role in research and development for the Fast Breeder Reactor (FBR) cycle in Japan, has a plan to procure Mixed Oxide (MOX) powder from the Rokkasho Reprocessing Plant (RRP), which the Japan Nuclear Fuel Limited (JNFL) is constructing at Rokkasho-mura, Kamikita-gun of Aomori prefecture for completion in July 2006. The MOX will be for fuel fabrication for the experimental ''JOYO'' and the prototype ''MONJU'' FBRs. The mixed oxide storage canister used in RRP is larger than that being used in the Tokai Reprocessing Plant and the Plutonium Fuel Production Facility (PFPF) in Tokai Works of JNC, and it contains approx. 36 kg of MOX powder. Because the existing packagings can not accommodate the RRP type of canister, the design of a new type packaging that can accommodate this canister was implemented. The structure of the packaging was arranged and the safety analysis of the package was carried out

  9. Unplanned shutdown frequency prediction of FBR Monju using fault tree analysis method

    In order to evaluate the operational reliability of prototype fast breeder reactor (FBR) MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using fault tree analysis (FTA) technique for the plant system model. The targeted devices are the following: primary heat transport system (PHTS), secondary heat transport system (SHTS), water and steam system (WS), plant protection system (PPS) and plant control system (PCS). In this paper was estimated the frequency of automatic reactor trips by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these systems. The analyses predicted 1.2/reactor year (RY) the value of unplanned shut down frequency by the internal factor of the system. The largest contributed event was function failure of SHTS accounting for 43% of total events followed by PHTS with 40%. The contribution factor of WS was only 4%. (author)

  10. Development of 3-D detailed FBR core calculation method based on method of characteristics

    A new detailed 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed in hexagonal-z geometry by combining the method of characteristics (MOC) and the nodal transport method. From the nodal transport calculation which uses assembly homogenized cross sections, the axial leakage is calculated, and it is used for the MOC calculation which treats the heterogeneity of fuel assemblies. Series of homogeneous MOC calculations which use assembly homogeneous cross sections are carried out to obtain effective cross sections, which preserve assembly reaction rates. This effective cross sections are again used in the 3-dimensional nodal transport calculation. The numerical calculations have been performed to verify 3-dimensional radial calculations of FBR (fast breeder reactor) assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region. (authors)

  11. Experience with, and programme of, FBR and HWR development in Japan

    Nuclear power generation in Japan is moving forward on the long-term development programme of nuclear power from the LWR to the FBR, essentially in the same way as in other advanced nuclear countries. In this development programme the unique HWR is also included; it can use plutonium produced in LWRs together with depleted uranium before the introduction of commercial FBRs. This report describes the status of the FBR and HWR development project being carried out by the Power Reactor and Nuclear Fuel Development Corporation (PNC) based upon the Long-Term Programme on Research, Development and Utilization of Nuclear Energy in Japan. Operational experience and technical results are shown for the experimental fast reactor JOYO (100 MW(th)), which reached initial criticality in 1977. The status of the 280 MW(e) prototype reactor MONJU, under construction as of 1982, is described. The conceptual design of the subsequent 1000 MW(e) demonstration plant is outlined, as is additional future planning. Research and development results, mainly carried out at Oarai Engineering Center of PNC, are shown. The 165 MW(e) prototype FUGEN is a heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor which uses plutonium mixed-oxide fuel. This report describes the relationship of the fuel cycle to the HWR in Japan and also discusses the operational experience of the prototype FUGEN, which has operated since 1979. Also described is the design of the 600 MW(e) demonstration plant and the programme of related research and development. (author)

  12. Analysis of the Bundle Duct Interaction using the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    PNC has been developing a computer code 'BAMBOO' to analyze the wire spaced FBR fuel pin bundle deformation under the BDI (Bundle Duct Interaction) condition by means of the three dimensional F.E.M. This code analyzes fuel pins' bowing and oval deformations which are dominant deformation behaviors of the fuel pin bundle under the BDI condition. In this study the 'BAMBOO' code is validated on the out-of-pile compression test of the FBR bundle (compression test) by comparing the results of the code analysis with the compression test results, and the highly irradiated (≥2.1x1027 n/m2, E > 0.1 MeV) bundle deformation behaviors are investigated from the viewpoint of the similarity to those in the compression test based on the analytical results of the code. (1) The calculated pin-to-duct minimum clearances as a function of the BDI levels in the compression test analysis agree with the experimental values evaluated from the CT image analysis of the bundle cross-section in the compression test within ±0.2 mm. And the calculated values of the fuel pins' oval deformations agree with the experimental values based on the pin diameter measurements done after the compression test within ±0.05 mm. (2) By comparing the irradiation induced bundle deformation with the bundle deformation in the compression test based on the code analysis, it is confirmed that the changes of the pin-to-duct minimum clearances with the BDI levels show equivalent trends between the both bundle deformations. And in this code analysis of the irradiation induced bundle deformation, contact loads between the fuel pins and the pacer wires are extremely small (below 10 kgf) even at about 3 dw of the BDI level compared to those in the compression test analysis. (J.P.N.)

  13. Survey report on trends of technical development on FBR cycle in Russia. Result report of business entrusted by the Japan Nuclear Cycle Development Institute

    This survey was carried out for aims to smoothly promote the FBR cycle cooperation carried out between the Japan Nuclear Cycle Development Institute and Nuclear Energy Agency in Russia Republic and to contribute their future cooperative planning, on a survey of technical developmental trend for items shown as follows: 1) recent trend on Russian FBR cycle technology, 2) Russian laws related on Russian FBR cycle cooperation, and 3) trends on separation and reprocessing technologies in Russia. Here was described on results on the survey, shown in the following items: 1) performing method, 2) Russian FBR fuel cycle; recent technical development, 3) basic laws on nuclear energy application in Russia, and 4) trends on separation and reprocessing technologies in Russia. (G.K.)

  14. Current status of research on FBR fuel behavior under accident conditions and the relevant NSRR program plan

    In the situation that the development of demonstration FBR is being materialized, a substantial research on safety of core fuels under accident conditions is required as the part of the research and development program. The experimental study of fuel integrity against over power accidents etc. and failure behavior is important to establish a criteria for safety evaluation of FBR's. In this report, the scope of the program which is planned in NSRR is shown after reviewing other related experiments and examining the research region left undone. Major in-core experiments on fuel failure are surveyed wide in the view point of experimental region and the inquired results are summarized. Subsequently, the items and methods due to the NSRR experiment program is discussed. The experimental facility plan and the results of preliminary analysis on the fuel energy deposition and temperature behavior are also introduced. (author)

  15. Consideration on rationalization of reactor safety systems. Reliability assessment of decay heat removal systems in commercialized sodium cooled FBR concepts

    For commercialization of FBR (Fast Breeder Reactor), the reactor safety systems are needed not only to have necessary and enough reliability but also to decrease the amount of materials in order that the FBR has economical competitiveness against LWR (Light Water Reactor) and another electrical power supply systems. In this study, reliability assessment, which calculates the occurrence frequencies of PLOHS (Protected Loss Of Heat Sink) sequences, was performed for three kinds of large size sodium cooled fast reactors with decreased number of loops and of support systems examined in the Feasibility Studies on Commercialized FBR System. The realistic evaluation was performed using the failure rate data of components based on the domestic LWR operating experience. The result is: All of the three kinds of commercialized FBR concepts are expected to achieve the frequencies of PLOHS sequences caused by internal events of the plant under 10-6/ry, assuming that common-mode-failure is excluded. In addition, the dominating cause of the coincidence of the incidents and the information that improves reliability of decay heat removal systems are summarized for each concept. In order to evaluate design margin, the reliability assessment was performed in the case that the capacity of natural circulation cooling was rein forced from 100%/3 x 3 loops to 50% x 3 loops or from 25% x 4 loops to 100%/3 x 4 loops easing to succeed decay heat removal. In that case, it is confirmed that the frequencies of PLOHS sequences decrease by about one order of magnitude. (author)

  16. Experimental and analytical studies on the multi-surface sloshing characteristics of a top entry loop type FBR

    Sloshing tests were conducted on the 1/8 scale model of the top entry loop type FBR with multiple free liquid surfaces, which are connected to each other by inverted U-shaped pipings. The sloshing characteristics of the top entry loop type reactor was established by these tests. Applicability of the analytical code and effectiveness of the anti-sloshing devides were also evaluated. (author)

  17. Sodium Exposure Tests on Limestone Concrete Used as Sacrificial Protection Layer in FBR

    Hot sodium coming in contact with structural concrete in case of sodium leak in FBR system cause damage as a result of thermo-chemical attack by burning sodium. In addition, release of free and bound water from concrete leads to generation of hydrogen gas, which is explosive in nature. Hence limestone concrete, as sacrificial layer on the structural concrete in FBR, needs to be qualified. Four concrete blocks of dimension 600 mm x 600 mm x 300 mm with 300 mm x 300 mm x 150 mm cavity were cast and subjected to controlled sodium exposure tests. They have composition of ordinary portland cement, water, fine and coarse aggregate of limestone in the ratio of 1: 0.58: 2.547: 3.817. These blocks were subjected to preliminary inspection by ultrasonic pulse velocity technique and rebound hammer tests. Each block was exposed for 30 minutes to about 12 kg of liquid sodium (∼ 120 mm liquid column) at 550 deg. C in open air, after which sodium was sucked back from the cavity of the concrete block into a sodium tank. On-line temperature monitoring was carried out at strategic locations of sodium pool and concrete block. After removing sodium from the cavity and cleaning the surfaces, rebound hammer testing was carried out on each concrete block at the same locations where data were taken earlier at pre-exposed stage. The statistical analysis of rebound hammer data revealed that one of the concrete block alone has undergone damage to the extent of 16%. The loss of mass occurred for all the four blocks varied from 0.6 to 2.4% due to release of water during the test duration. Chemical analysis of sodium in concrete samples collected from cavity floor of each block helped in generation of depth profiles of sodium monoxide concentration for each block. From this it is concluded that a bulk penetration of sodium up to 30 mm depth has taken place. However it was also observed that at few local spots, sodium penetrated into concrete up to 50 mm. Cylindrical core samples of 50 mm x 150

  18. Electronic manual of the nuclear characteristics analysis code-set for FBR

    Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

  19. Assessment of fuel coolant interactions (FCIs) in the FBR core disruptive accident (CDA)

    Application of general behavior principles (GBPs) and consideration of relevant contact modes suggest that only incoherent small-scale fuel coolant interactions (FCIs) with negligible damage potential appear possible with the molten oxide fuel-liquid sodium system as the fuel disperses away from the core into a coolable non-critical array. In contrast to the SPERT-1, BORAX-1 and SL-1 nuclear transients that ultimately led to energetic vapor or steam explosions, the presence of molten fuel and liquid sodium in the FBR core always requires the presence of solid cladding which separates the fuel and coolant and, hence prevents energetic FCIs prior to coolant escape. Furthermore, unlike the CORECT-II experiments which examined dynamic re-entry of liquid sodium on molten fuel pools that resulted in unstable interfaces leading to significant sodium entrapment and relatively energetic FCIs, the prevailing contact mode in the FBR core disruptive accident (CDA) scenario is displacement of the lighter and less viscous liquid sodium by the heavier and more viscous molten fuel resulting in stable interfaces with no significant sodium entrapment and FCIs. Dynamic re-entry of liquid sodium into the core is not possible with the two-component steel vapor-liquid sodium system, since the interface contact temperature upon steel vapor condensation is well in excess of the sodium boiling temperature. A pressure reduction in the steel vapor region due to condensation is immediately compensated for by an equivalent pressure increase due to sodium evaporation. Finally, considering that the molten oxide fuel-liquid sodium interface contact temperature is well below the sodium homogeneous nucleation temperature which in turn is well below the fuel melting temperature, not only eliminates the potential for large-scale vapor explosions as molten fuel streams are injected into liquid sodium pools, but also implies that small scale superheat explosions are possible which are consistent with

  20. Material properties of high Cr-Mo steel. 3. Mechanical properties of HCM12A (FBR) after thermal aging

    In the FBR components, cyclic thermal loads are predominant and creep deformation must be taken into account. Therefore, the applicability of high chromium ferritic steel for structural material of the future advanced Fast Breeder Reactor is investigated in the feasibility study on commercialized FBR cycle systems, since both of thermal properties and high temperature strength of the steel are superior to those of conventional austenitic stainless steels. In this study, tensile, hardness, impact and relaxation tests are conducted in order to evaluate the basic mechanical properties of each HCM12A(2001-FBR). The material is aged at 600degC for 3000h and 6000h. The aged materials are also tested as well as the as-received one. As a result, the following conclusions are obtained; (1) Though 0.2% proof stress and ultimate tensile strength of thermal aged specimens are smaller than those of as-received ones, those properties still larger than those of as-received Mod.9Cr-1Mo and HCM12A steels for thermal power generating plants and the values of Sy and Su which were given in a tentative plan of HCM12A material properties guideline. (2) The ductility of the material are degraded by aging. The Fracture elongation and reduction of area of the as-received are inferior to those of Mod.9Cr-1Mo and HCM12A steels. (3) Upper shelf energy (USE) of the materials in Charpy impact test decreases with aging. 600degC-6000h aging process degrades the USE of HCM12A(2001-FBR) from about 110 J/cm2 to 65 J/cm2. The USE of the HCM12A(2001-FBR) steel is less than 1/2 and 2/3 of those of Mod.9Cr-1Mo and HCM12A steels, respectively. (4) Relaxation stress material aged at 600degC for 6000h was smaller than as-received one in both 0.15% and 0.30% strain controlled stress relaxation test. (5) The metallurgical examinations suggest that coarsening of M23C6 and precipitation of Laves phase in prior γ-grain, lath and packet boundaries due to thermal aging result in ductility decrease. (author)

  1. System performance test of mechanical decladding system for FBR fuel reprocessing

    Japan Nuclear Cycle Development Institute (JNC) has been developing a new mechanical decladding system for FBR oxide fuel. The mechanical decladding system consists of the fuel pin crushing step (mechanical crusher), the hull separation step (magnetic separator) and the hull rinsing step (melting separator). The system mock-up test device is fabricated, and the performance tests are carried out with simulated oxide fuel pins. As the results, the following performance of each device was confirmed. The lifetime of the crushing blades of mechanical crusher were more than 1 year, and practical use. The vibration classifier could be separated 3 groups (wrapping wire, small particles and large particles), and recovered the majority of the wrapping wire. High fuel purity and a high fuel separation efficiency were obtained by the multi-stage magnetic separator. The fuel and the hull could be roughly separated, and the majority of the fuel could be recovered easily by the Cold Crucible Levitation Melting (CCLM) technology. As the above results, each device performance was satisfied the specification of commercialized scale reprocessing plant, and the high fuel recovery performance achieved by the combination of the multi-stage magnetic separation and the hull melting separation. Therefore, we can establish innovative engineering design of the mechanical decladding system. In the future, hot demonstration data using spent fuel will be carried out, and remote maintenance design will be conducted. (author)

  2. The development and application of overheating failure model of FBR steam generator tubes

    The following items have been studied to evaluate overheating failure of FBR steam generator heat transfer tubes: 1) To establish a structural integrity analysis method, 2) To improve and validate blow down analytical method, 3) To quantitatively validate the entire overheating analysis model by sodium water reaction data. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantitatively shown through the analysis: 1. The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. 2. Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. 3. Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin. (J.P.N.)

  3. Enhancing the performance of FBR fuel and core structural materials through PIE

    Post irradiation examination (PIE) aims at generating and evaluating data that provide valuable insight into the in-reactor behavior of fuels and core structural materials. PIE essentially has three major roles namely (i) life assessment and extension of burnup (ii) fuel and structural material development and (iii) failure analysis and is a vital link in the fuel cycle. The challenging task of setting up a hot cell facility for PIE of Fast Breeder Test Reactor (FBTR) was accomplished successfully and irradiation performance of the FBTR mixed carbide fuel was assessed stage wise at various burnups starting from 25 GWd/t upto 155 GWd/t. With FBTR being used as a test bed for irradiation experiments on various FBR fuels and structural materials, PIE of various materials subjected to experimental irradiation like the PFBR MOX fuel, FBTR grid plate material have also been carried out to provide valuable feedback to the designers. This paper will discuss the performance assessment of the FBTR driver fuel and structural material at different burn-up, and the role of PIE in progressive enhancement of the burn-up to a record high of 165 GWd/t. The role of PIE in performance assessment of materials like B4C control rod, MOX fuel of PFBR composition and FBTR grid plate material will also be presented

  4. Method of start-up of rotary plug sealing devices in FBR type reactors

    Purpose: To rapidly and safely start-up the rotary plug sealing device by controlling to eliminate the pressure difference in the pressures of gases exerting on the liquid surfaces in the inner and the outer cylinders of a sealing alloy vessel in the rotary plug of a FBR type reactor. Method: In a case where an abnormal state results in the pressure difference of gases exerted on the liquid surfaces in the inner and the outer cylinders of a vessel charged with sealing alloy in a rotary plug and the sealing valve for the back-up gas supply tube is rapidly closed to seal the sealing portion, the pressure in the gas supply tube is controlled so that the pressure difference in the gases exerted on the liquid surfaces in the inner and outer cylinders while closing the sealing valve. Then, after conforming that the pressure is controlled to a predetermined level at which the pressure difference can be regarded to be zero, the sealing valve is gradually opened while regulating the pressure in the gas supply tube so as to maintain the pressure difference to a predetermined level. This prevents the occurrence of external disturbances upon opening of the sealing valve and enables rapid and safety start-up for the rotary plug sealing device. (Moriyama, K.)

  5. Component analysis of sodium void reactivity of step type FBR cores with group-wise Monte Carlo code 'GMVP'

    Reactivity components composing the sodium void reactivity in a FBR core are analyzed by group-wise Monte Carlo Code GMVP, which has been developed by JAEA. The typical way to analyze the reactivity components is to use the perturbation method based on the diffusion calculations, while the diffusion approximation cannot be appropriately applied to some types of FBR cores containing large cavity regions. But, in order to prospect the optimized FBR core with negative sodium void reactivity, we need to the components of the sodium void reactivity of cores which have a small void reactivity, which cores are sometimes accompanied with adjacent large cavity regions or gas plenum zones. In this study, we have employed GMVP to simulate the cavity region exactly in geometry and to evaluate the neutron behavior rigorously in reactor physics. The cross section library used is JFS-3-J3.3 70 group constant set that is complied from JENDL-3.3 library. The objective core is a 'step type' two zone core, which has a lower inner core height relative to the height of the outer core, and the upper axial blanket is eliminated to enhance the neutron leakage in the upper ward at void conditions. The reactivity component by neutron leakage is derived from the difference of the k-effective of direct calculation of GMVP between the intact and void cores, and that of non-leakage components evaluated by using real and adjoint flux that are calculated with GMVP. In the paper, the change of the contributions of the both components is presented when the core height is changed along with the void reactivity of the cores. (author)

  6. The development and application of overheating failure model of FBR steam generator tubes. 2

    The JNC technical report 'The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes' summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. 1. On the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. 2. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. 3. Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure. (author)

  7. Burnup behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel cycle economy

    to established the same burnup performance, mainly because of poisonous influence of 236U. The problem of 236U is, rather than its neutron absorptivity, its transformation to 237Np having a large cross-section area of neutron capture. The MOX fuels manufactured with the matrixes sourced in the recovery uranium, however, can compensate the neutron toxicity of 236U with 235U enriched in the matrixes while contributing to core burnup. So, these MOX fuels require a decrease of plutonium, compared with the MOX fuel made with the depleted natural uranium. The quantitative fuel reproduction in the first recycle was estimated as follows: In the case of A-type matrix, the MOX fuels for 13.7 PWRs and the reformed uranium fuels for 12.4 PWRs can be obtained from the conventional spent fuels of 100 PWRs. In the case of B-type matrix, the MOX fuels for 13.2 PWRs and the reformed fuels for 14.2 PWRs can be manufactured from the same spent fuels. In the case of C-type matrix, the MOX fuels and the reformed uranium fuels can be provided for 12.7 and 14.2 PWRs, respectively. Therefore, the fuel recycle efficiencies in the PWR system are 26.1%, 27.4 % and 26.9 % for the cases A, B and C, respectively. (partially reported in 4th RRTD Int. Workshop for Asian Nucl. Prospect (1st Asian Nuclear Prospect Workshop), October 19-21, 2008 Kobe, Japan.) The multi-recycle in the light water reactor (LWR) fuel system, however, will be involved by that 236U in reformed fuels increases more and more with recycle times. The spent-fuel resources are recommended to be utilized in the fast breeder reactor (FBR) fuel cycle. Thus, we have been studying the effective use of uranium and plutonium resources in the FBR cycle cooperating with the LWR system. In this work, a FBR of practical type (1.5 GWe-class) was conceptually designed in the SRAC numerical system, referring to a commercial-type FBR model proposed by Japan Atomic Energy Agency and The Japan Atomic Power Co. Its main design parameters are

  8. Human development in Japan and abroad using the prototype FBR 'Monju' towards the next-generation age

    Japan is engaged in research and development for an innovative FBR, targeting commercial operation to start around 2025 via the FaCT (Fast Reactor Cycle Technology Development) project. To prepare for the new FBR age, INITC has been working on human resource development using Monju towards the next-generation age not only for Japan, including companies and students, but also for the world aiming at becoming a base of the international educational training. Effective mastering technology requires educational training harmonized through lectures and exercises, consequently, providing such education needs preparing substantial hardware and developing software. As for the preparation of hardware, the Fast Reactor Training Facility, which consists of two training facilities for systematically teaching the technologies for sodium handling and maintenance related to FBR technology, was built in May 2000, based on the lessons learned from the December 1995 sodium leak accident. As regard in to operator training, the hardware of MARS (Monju Advanced Reactor Simulator) was reconstructed corresponding to the sodium leak safety measures, and simulator function was also improved in order to advance operator training. The following advances concerning the software were also carried out based on the teachings obtained from the leak accident as well as hardware: a) Strengthening operator training framework by establishing educational training guidance, improving operation manuals and introducing an operator training evaluation system (SAT: Systematic Approach Training); b) improving training curriculums related to plant system engineering and sodium handling technologies, which cover teaching from the basis to the expert levels. A total of 26 training courses were established, such as 8 simulator courses, 5 FBR plant system engineering courses, 6 sodium handling courses and 7 maintenance courses. And each course has been continuing even while Monju operation has been stopped for

  9. 3-D numerical simulation on the vibration of liquid sodium's free surface in sodium pool of FBR

    This paper succeeds in simulating three-dimensional incompressible flows with free surface, complicated in-flow and out-flow boundary conditions and internal obstacles, and also can treat these fluid flows in arbitrary shape vessel using a partial cell. According to all kinds of the element influencing the free surface's vibration in sodium pool it may give the various wave's form, the highest and lowest position, and the amount of the vibration. This paper introduces the brief principle of VOF numerical method, develops the computational program based on NASA-VOF3D, provides some results about the free surface's vibration in sodium pool of FBR

  10. Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

    Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks

  11. Plans of Verification Tests for the ACTOR Code Analyzing Fission Products Behavior in Primary Heat Transportation System of FBR

    The ACTOR is the analysis code to calculate the transfer behavior of fission products (FPs), which are released from the fuel plenum or fuel pellets due to a cladding breach caused by temperature increase during an accident, in the coolant sodium and cover gas of a fast breeder reactor (FBR) plant. Principal analysis models adopted in the ACTOR code such as 'FP release model', 'Bubble transfer model in sodium', 'FP transfer model from bubble to sodium' and 'FP adsorption model' are constructed and verified based on the experimental results relating to each phenomenon. But, some of the analysis models are still desirable to be verified by experiments because those have significant influence on the evaluation results. In this paper, principal analysis results of FPs' behavior in the primary heat transportation system (PHTS) during PLOHS event of FBR plant which are obtained by using ACTOR code are introduced to focus the analysis models to be verified, and the plans of verification tests for the ACTOR code are described. (author)

  12. Application of wavelet analysis to signal processing for Eddy--Current Testing of FBR steam generator tubes

    We have been studying an advanced signal processing method for ECT (Eddy-Current Testing) to detect flaws in heat exchanger tubes of steam generators in a FBR (Fast Breeder Reactor) plant. The data obtained from the testing of these steam generator tubes showed some noise characteristics different from ones in light water reactor plants. Thus, an improvement of the signal-to-noise ratio of the processed ECT data is desired in order to increase reliability of the inspection. We have evaluated the signal-to-noise ratio as a performance measure using the actual plant data by means of wavelet analysis technique, which has been applied successfully in many fields to detect anomalous signals. First, we synthesized an ECT data containing flaws by combining the noisy signals from an in-field inspection of one heat exchanger tube with separately measured flaw signals from a calibration piece that had various machined flaws. Next, we evaluated the performance of the wavelet analysis technique by comparing it with a reference Fourier method. The performance of this technique is much better than by conventional analysis. For example, the signal-to-noise ratio of the discrete wavelet analysis method is 1.4 times greater than that by a conventional, low-pass filter method. Thus, we have shown that wavelet analysis has advantages over conventional method of detecting flaws in FBR steam generator tubes because of its sensitivity to localized events, such as sudden signal changes. (author)

  13. Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

    Joo, Young Sang; Kim, Jong Bum; Kim, Seok Hun; Yoo, Bong

    2001-03-01

    Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks.

  14. Transport criticality analysis for FBR MONJU initial critical core in whole core simulation by NSHEX and GMVP

    FBR MONJU Initial Critical Core (ICC) criticality problem has been solved by deterministic and Monte Carlo transport methods by the codes NSHEX and GMVP. The analysis has been carried out in different energy-groups approximations. As a result the effect of cross-section (XS) condensation from 70 into few energy-group structures by different collapsing methods has been evaluated. The 3D discrete-ordinate code NSHEX has been applied for wide range of core simulations-from whole core, considering the fissile, fertile and shielding regions to simplified models that simulate an increased neutron leakage. It has been found that there is room for improvement in the assessment of the neutron leakage in the few energy-group approximations. The good agreement between NSHEX and GMVP results, especially without XS collapsing, is pointed out as a conformation for the applicability of the code NSHEX in FBR 3D whole core calculations. Some practical conclusions have been extracted that are important for the implementation of the code NSHEX in the standard criticality analysis. (author)

  15. Engineering scale tests of mechanical disassembly and short stroke shearing systems for FBR fuel assembly

    Japan Atomic Energy Agency (JAEA) and The Japan Atomic Power Company (JAPC) have been developing an advanced head-end process based on mechanical disassembly and short stroke shearing systems as a part of Fast Reactor Cycle Technology Development (FaCT). Fuel pins for a fast reactor are installed within a hexagonal shaped wrapper tube made of stainless steel. In order to reprocess the fast reactor fuel pins, they must be removed from the wrapper tube and transported to the shearing system without failure. In addition, the advanced aqueous reprocessing process, called 'NEXT' (New Extraction System for TRU Recovery) process requires a solution of the spent fuel with relatively high concentration (500g/L). JAEA and JAPC have developed the mechanical disassembly and the short stroke shearing technology which is expected to make fragmented fuel to satisfy these requirements. This paper reports the results of engineering scale tests on the mechanical disassembly and short stroke shearing systems. These tests were carried out with simulated FBR fuel assembly and removed pins. The mechanical cutting method has been developed to avoid fuel pin failure during disassembly operation. The cutting process is divided into two modes, so called 'slit-cut' for cutting the wrapper tube and 'crop-cut' for the end plug region of the fuel pin bundle. In the slit-cut mode, the depth of cutting was automatically controlled based on the calculated wastage of the cutting tool and deformation of the wrapper tube which had been measured before the cutting. This procedure was confirmed to minimize the fuel pin failure which was hard to prevent in the case of laser cutting. The cutting speed was also controlled automatically by the electric current of the cutting motor to lower the load of the cutting tool. The removed fuel pins were transported to the shearing machine, whose fuel shearing magazine width was set to be narrow to realize the suitable configuration for the short stroke shearing

  16. Analysis of core physics test data and sodium void reactivity worth calculation for MONJU core with ARCADIAN-FBR computer code system

    In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)

  17. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P0/s (P0: steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

  18. Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR

    JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)

  19. Photo-stimulated luminescence of calcium co-doped BaFBr : Eu2+ x-ray storage phosphors

    The influence of calcium co-doping on the optical properties of the x-ray storage phosphor BaFBr : Eu2+ is determined by photo-stimulated luminescence techniques. It is found that the incorporation of calcium into the lattice results in a broadening of the photo-stimulation peak due to a calcium induced FA(Br-, Ca2+)-centre with stimulation maxima at 540 and 680 nm. The optical cross-sections for the photo-stimulated process are determined by utilizing stimulation light with linearly increasing intensity. Furthermore, it is shown that the sensitivity for x-rays, i.e. the number of storage centres formed during x-ray exposure, increases up to a doping level of 1 mol% while it drops rapidly at higher calcium concentrations

  20. Evaluation of scatter radiation in digital radiological condition by using phostimulated luminescence (BaFBr:Eu{sup 2+})

    Min, Jung Whan [Dept. of Radiological Science, Shin-Gu University, Sungnam (Korea, Republic of); Han, Seong Gyu [Dept. of Bio-convergence engineering, Graduate school, Korea University, Seoul (Korea, Republic of); Kim, Jung Min [Dept. of Radiological Science, Korea University, Seoul (Korea, Republic of); Lee, Joo Ah [Dept. of Oncology, Catholic University of Korea Incheon St.Mary,s Hospital, Incheon (Korea, Republic of); Kim, Ki Won [Dept. of Radiology, Samsung Medical Center, Seoul (Korea, Republic of); Jeong, Hoi Woun [Dept. of Radiological Science, Beakseok Culture University, Cheonan (Korea, Republic of)

    2014-06-15

    The purpose of this study is evaluated scatter radiation in digital radiological condition by using photostimulated luminescence (BaFBr:Eu2+). Experiment condition changed kVp (from 50 kVp to 120 kVp), filed size (from 4 × 4 cm{sup 2} to 26 × 26 cm{sup 2}) and phantom thickness (from 1 cm to 15 cm). This method was analysed ImageJ and characteristic curve of CR. This results was scatter radiation to primary radiation ratio increased from 50 kVp to 70 kVp, and it was fixed at over 80 kVp. The scatter radiation to primary radiation ratio are increased according to increasing the ratio of field size. Scatter radiation is also increased by increasing the phantom thickness.

  1. Evaluation of scatter radiation in digital radiological condition by using phostimulated luminescence (BaFBr:Eu2+)

    The purpose of this study is evaluated scatter radiation in digital radiological condition by using photostimulated luminescence (BaFBr:Eu2+). Experiment condition changed kVp (from 50 kVp to 120 kVp), filed size (from 4 × 4 cm2 to 26 × 26 cm2) and phantom thickness (from 1 cm to 15 cm). This method was analysed ImageJ and characteristic curve of CR. This results was scatter radiation to primary radiation ratio increased from 50 kVp to 70 kVp, and it was fixed at over 80 kVp. The scatter radiation to primary radiation ratio are increased according to increasing the ratio of field size. Scatter radiation is also increased by increasing the phantom thickness

  2. Analysis of mass oscillation of a multiple free-surface FBR with a non-linear lumped mass model

    The primary circuit of a Japanese FBR (fast breeder reactor) is composed of a reactor vessel, three pump vessels and three IHX (intermediate heat exchanger) vessels, which are connected with top-entry piping system. Since the sodium has several free surfaces contacting argon cover gas in each vessel, the primary circuit forms a multiple free-surface system. The behavior of sodium elevation in each vessel is not straight forward, because the cover gas volumes in the vessels are connected with gas tubes, consequently enhancing the complexity of the system. In the present study, a lumped mass model was developed in order to examine the effects of the non-linearity and cover gas on the oscillatory behavior. The non-linearity of the system was incorporated into the basic equation by using the Lagrange equation, while gas dynamics equations were derived, assuming state of gas obeys polytropic law. The nonlinear model was verified with a 1/8-scale model test conducted by JAPCO, where water and air were used as simulant fluids for sodium and argon gas. Comparison between the numerical results of the mathematical model and the experimental results of the reduced-scale test was made for both reactor trip event and gas-line rupture event. The quantitative agreement of temporal change of water level justified the validity of the present model. The results of eigen-value analysis for the basic equation show that there are two eigen modes for manometer mass oscillation. One is the mode where all IHX sodium levels oscillate in in-phase motion, while the other is the mode where the sodium level in a reactor vessel stays still. The periods of both modes strongly depend on existence of cover gas and the period decreases as the pressure increases. The present model was applied to several events presumable in an FBR to demonstrate the applicability of the code. For example, the computed results show that the mass oscillation behavior is very sensitive to the state of cover gas lines

  3. Experimental and analytical studies on the multi-surface sloshing characteristics of a top entry loop type FBR

    For a system with multiple vessels which have a free liquid surface and are connected with inverted U-shaped piping, it is important to establish a highly accurate technique of evaluating the sloshing behavior that occurs in the multiple liquid surface system when an earthquake strikes.In the present study, a series of vibration tests were conducted to clarify the basic characteristics of sloshing using parameters such as the in-vessel structure, multi-surface connection, in-vessel flow, and so forth. For testing, a 1/8 scale model of the top entry loop type FBR primary loop was used. Further, the applicability of the analytical code was evaluated, and the code was found to estimate the mode shape and natural frequency of sloshing with high accuracy. (orig.)

  4. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    Japan Nuclear Energy Safety organization (JNES) is developing severe accident analysis codes in order to apply to the probabilistic safety assessment (PSA) for a typical fast breeder reactor (FBR). The AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary and the release fraction to the environment of fission products (FP). This report summarized results analyzed using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass (CVBP) scenario, and the containment failure scenario due to hydrogen deflagration or detonation. The results showed that the coolant temperature of the primary system and the secondary system in the PLOHS sequence increased at the almost same temperature, and the creep damage to the reactor coolant boundary became significant when coolant temperature exceeded about 1,100 K. The release fractions of FP in the CVBP case were estimated to be 0.99 for Xe, 0.14 for iodine, 0.44 for Cs and 0.01 for non-volatile tetravalent Ce. The release fractions of FP in the containment vessel failure case due to hydrogen burning were estimated to be 0.82 for Xe, 0.06 for iodine, 0.06 for Cs and 0.003 for non-volatile tetravalent Ce. In the present study, release fractions of FPs to the environment were obtained for the CVBP and the containment failure cases of the PLOHS accident sequence for the typical FBR plant. (author)

  5. Burn-up behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel economy

    Considering the strategic security of uranium resources, the authors investigated the effectual use of fuels recycled in a current LWR system (1.1 GWe-class PWRs of standard type), where they supposed the uranium fuel remade by re-enrichment of the recovery uranium from PWR spent fuels and the MOX fuel produced with the recovery plutonium as main fissile, and examined three cases of MOX fuels prepared by making choice of these matrixes among (A) the recovery uranium, (B) the depleted uranium as waste from natural uranium enrichment and (C) the depleted one sourced from recovery uranium re-enrichment. The result suggests that the multi-recycle of fuels in the LWR system brings the decline in fuel qualities. Particularly, the re-enrichment of recovery uranium brings the issue of an increase of 236U in remanufactured fuel. Thus, in order to investigate the burn-up of fast reactor fuels sourced from the PWR fuel system, they designed a model core of practical-FBR with reference to a concept of FBR reported by JAEA, using the SRAC numerical system. The burn-up behavior of FBR fuels was analyzed which were sourced in the original uranium spent-fuel and the remade uranium spent-fuel. And also, the breeding behavior of blanket materials was investigated which were individually of the depleted natural uranium, the recovery uranium from the original uranium spent-fuel and the recovery uranium from the remade uranium spent-fuel. The fissile 235U in FBR fuels reduces the burden of plutonium while the containment of 236U declines the neutron multiplication in FBRs. (author)

  6. Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

    A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

  7. Bioethanol fermentation by recombinant E. coli FBR5 and its robust mutant FBHW using hot-water wood extract hydrolyzate as substrate.

    Liu, Tingjun; Lin, Lu; Sun, Zhijie; Hu, Ruofei; Liu, Shijie

    2010-01-01

    Hemicellulose is a potential by-product currently under-utilized in the papermaking industry. It is a hetero-carbohydrate polymer. For hardwood hemicelluloses, D-xylose is the major component upon depolymerization. At SUNY-ESF, wood extracts were obtained by extracting sugar maple wood chips with hot water at an elevated temperature. The wood extracts were then concentrated and acid hydrolyzed. Ethanologenic bacteria, E. coli FBR5, had a good performance in pure xylose medium for ethanol production. However, FBR5 was strongly inhibited in dilute sulfuric acid hydrolyzate of hot-water wood extract. FBR5 was challenged by hot-water wood extract hydrolyzate in this study. After repeated strain adaptation, an improved strain: E. coli FBHW was obtained. Fermentation experiments indicated that FBHW was resistant to the toxicity of hydrolyzate in the fermentation media of concentrated hydrolyzate, and xylose was completely utilized by the strain to produce ethanol. FBHW was grown in the concentrated hydrolyzate without any detoxification treatment and has yielded 36.8g/L ethanol. PMID:20478373

  8. Recycling of actinides produced in LWR and FBR fuel cycles by applying pyrometallurgical process

    Integrated pyrometallurgical technology will be applied on the fuel reprocessing of oxide and metal fuels and on the recovery of transuranium elements. The main processes consisted of electrorefining, reductive extraction and waste treatment. The oxides will be reduced to metals by using reductant agent prior to the application of electrorefining. The high level liquid waste coming from purex type of reprocessing of LWR fuels can be also treated in order to separate transuranium elements at the reduction extraction process. The salt waste treatment was evaluated on the methods of direct solidification by artificial rock and vitrification after electrolysis. The process flow was proposed based on the experimental results for the partitioning of transuranium elements from high level liquid waste. (author)

  9. Development of ultrasonic two-dimensional arrayed transducer for visual inspection under high temperature sodium in FBR

    An under sodium imaging technique has been developed by means of the synthetic aperture focusing technique (SAFT) using a single ultrasonic transducer with mechanical scanning. Mechanical scanning imaging, however, might have some practical difficulties such as shortening a data sampling time and so precise requirements a scanning mechanism. This paper describes a newly developed two-dimensional arrayed ultrasonic transducer (Undersodium multiple transducer; Multi-Transducer) which consists of piezoelectric ceramics for visual inspection under high temperature sodium in FBR. Effects of parameters such as number of piezoelectric ceramic tips, locational arrangements of tips and acoustic properties of materials, on imaging were studied by a computer simulation. Fundamental specifications of Multi-transducer were also examined and a conceptual design was obtained through some cases of computer simulation. As a result, about 400 piezoelectric ceramic tips might be needed for the full model of Multi-Transducer to satisfy a required resolution of imaging. A Multi-Transducer section model was designed and manufactured, which consisted of 25 piezoelectric ceramics, and acoustic experiment were made to confirm required performances under water and silicon oil (at 230degC). (author)

  10. In situ measurement of photo-stimulated luminescence from BaFBr:Eu2+ irradiated by He+ ion beams

    Radiation effects of photo-stimulated luminescence (PSL) from BaFBr:Eu2+ used for an imaging plate have been studied with various radiation sources in these years. In situ measurements of both afterglow and PSL emission from the imaging plate have been carried out with 1 MeV He+ ion beam irradiation in this report. Prompt luminescence and afterglow emission peaked at 400 nm in wavelength under and just after ion beam irradiation have been observed, and the PSL has been observed after the irradiation by the excitation with 500 or 600 nm light source. We have elucidated various characteristics of the PSL of the imaging plate by the ion beam irradiation as follows. (1) The emission just after the ion beam irradiation includes the PSL and afterglow emission. (2) The behavior of the time dependence of the PSL is different from that of the afterglow emission. Fading decay time of the PSL is longer than that of the afterglow emission. (3) The spectrum of the PSL is composed of more than four transition levels

  11. Review of reactor physics activities relevant to FBR and ATR programmes in PNC, Japan, June 1977 to October 1978

    Reactor physics works in Power Reactor and Nuclear Fuel Development Corporation (PNC) are carried out in support of the FBR and ATR development programmes. The works are in progress with high efficiency, in cooperation of other related organizations in Japan and also overseas. The experimental fast reactor ''Joyo'' reached its full power of 50 MWt, being operated now to obtain technical data and experiences. The prototype fast breeder reactor ''Monju'' is nearing completion of its final design, and the surveys on its proposed site have been finished. The advanced thermal reactor ''Fugen'', a heavy water-moderated and boiling light water-cooled, reactor reached the criticality, and the power generation was successfully achieved. The data to be obtained are used for the development of a large demonstration reactor. After describing on these types of reactors briefly, the works on reactor physics are described as follows: for fast reactors, mockup experiment and analysis, evaluation of actinide nuclear data, development of core analytical method, and research on shielding; and for ATR, research with a deuterium critical assembly. (Mori, K.)

  12. Measurements of relative power distributions in the axially simulated heterogeneous FBR cores by γ-scanning method

    Measurements of relative power distributions were made using the γ-scanning method in the partially simulated cores of the axially heterogeneous FBR in order to study power flattening by introducing the inner blanket at core midplane and power distortion by insertion of simulated B4C control rod in the core. Power peaking factor was decreased by about 12 % in FCA XII-1 assembly in comparison with FCA XI-1 assembly, and the value was 1.11 +- 1.4 %. Distortion in power distribution caused by introducing the simulated B4C control rod in the FCA XII-1 assembly was obtained from the measured power distributions and propagation distance of the distortion was examined. It was observed that the inner blanket played a role to cease the propagation of distortion from the upper to lower half assembly. Calculations were made for all cores. Calculated results predict the measured results fairly well in the core region and inner blanket. A large descripancy remains in the outer blanket. (author)

  13. Liquid metal storage tank

    The present invention concerns a liquid metal storage tank used for an FBR type reactor plant. It comprises a tank main body disposed in a pit chamber, a sealing tub disposed at an upper outer circumferential surface of the tank main body, a roof portion which closes the opening a the upper end of the pit chamber, a sealing partitioning cylinder suspended from the lower surface of the roof and having its lower end extended to the inside of the tub and a sealing liquid metal filled in the tub. The tank main body is kept at a high temperature by the liquid metal while the roof in the upper portion of the pit chamber is kept at a low temperature. Further, since the tank main body and the inside of the pit chamber are sealed by the sealing partitioning cylinder, no large thermal stresses are caused to the wall of the tank main body. Even if hydrogen gases are generated in the tank main body, since they can be released to the inside of the pit chamber, the integrity of the tank can be maintained, even if abrupt pressure elevation is caused in the tank main body. (I.S.)

  14. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    JNES is developing severe accident analysis codes in order to apply to the probability safety analysis (PSA) for a typical fast breeder reactor (FBR). AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary, and the discharge rate to the environment of fission products (FP). This report summarizes analysis results using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass scenario (CVBP) and the containment failure scenario by hydrogen deflagration or detonation. The coolant temperature of the primary system and the secondary system in the PLOHS sequence increases at the almost same temperature, and the creep damage to the reactor coolant boundary will become remarkable if coolant temperature exceeds about 1,100 K. In the CVBP scenario, when an intermediate heat exchanger is ruptured by creep and the boundary of the secondary system is failed, the path from the primary system to environment is formed. Then, the reactor vessel (RV) is failed and sodium in the primary coolant system releases into the reactor vessel room (RV room). Sodium of high temperature which fell in the RV room damages the floor liner, and generates hydrogen by a reaction with concrete. In addition the reactor core is exposed into atmosphere and the core temperature increases with decay heat and then volatile FP and non-volatile FP are released to the environment through the secondary system from the primary system. In the non-CVBP scenario which the intermediate heat exchanger does not fail by creep, core debris falls into the RV room after reactor vessel failure or evaporation of sodium coolant molten. FPs released from the reactor vessel are retained in the RV room, the primary system room, the containment dome and so on. The hydrogen generated by sodium-concrete reaction and

  15. Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

    The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

  16. Dose response of BaFBrI:Eu2+ storage phosphor plates exposed to megavoltage photon beams

    The BaFBrI:Eu2+ storage phosphor plate (SPP) is a reusable radiation image detector, widely used in diagnostic computed radiography, x-ray crystallography and radioactive tracer studies. When exposed to ionizing radiation, the SPP stores a latent image until it is scanned with a red reading laser which causes blue photostimulated luminescent (PSL) photons to be emitted. The mechanism of formation of the latent image is still poorly understood, especially for megavoltage photon beams. In order to gain insight into this mechanism and aid applications to high-energy beam dosimetry, the authors have directly determined the SPP generation efficiency, W, the energy required to produce one quantum of emitted PSL when it is irradiated by 60Co and 6 MV photon beams. This was done in four steps: 1. The SPP, in a water-equivalent plastic (WEP) phantom, was exposed to a 60Co or 6 MV beam, which had been calibrated to give a known absorbed dose to water in a water phantom at the position of the sensitive layer of the SPP. 2. Monte Carlo simulations were used to calculate the ratio of the dose to the sensitive layer in the WEP phantom to the dose to water at the same position in a water phantom. 3. A bleaching experiment was used to determine the number of photons emitted by a plate given a known dose. 4. The generation efficiency was calculated from the number of photons and the dose. This method is much more direct than previous calculations for kilovoltage x-ray beams based on quantum noise analysis. W was found, within experimental uncertainty, to be 190 eV for 60Co and 160 eV for 6 MV, independent of dose. The values for kilovoltage x-ray beams determined previously agree, within their large uncertainty, with these values for megavoltage beams

  17. An evaluation study on ULOF event sequences in the prototype FBR. An evaluation of CDA reflecting the latest knowledge

    The sequences of ULOF (unprotected loss-of-flow) event in the prototype FBR has been evaluated, as a part of the research and development (R and D) in the reactor safety research, reflecting the latest experimental and analytical knowledge on CDA (core disruptive accident) which has been accumulated at O-arai Engineering Center. In this study, an emphasis is placed on the event sequence of the melt progression phase ('transition phase') which has been recognized as one of the important issues of CDA analysis. The major parameters to be considered in this phase are the change of the mobile molten fuel mass and the history of the fuel motion, and also the relation between these parameters and energy generation mechanism. The following conclusions were deduced from the analysis and assessment. (1) The initial transient in ULOF accident sequence, which is driven by the void reactivity, does not produce a significant energy due to the existence of the inherent mechanism which mitigates the reactivity insertion. The disrupted fuel remains dispersed in the core and the initial transient is followed by the transition phase. (2) In the most probable event progress in the transition phase, the reactivity lowers continuously with the core melting propagation and the accident eventually terminates at the permanent sub-critical state without any significant energy release. (3) The recriticality event which produces kinetic energy is only possible under the conservative assumptions in which these mitigating mechanisms are restricted and the consistency of event progress is modified hypothetically. (J.P.N.) 55 refs

  18. Sodium removing facility for core-constitutional elements of FBR type reactor

    Reactor core-constitutional elements as spent reactor core fuel assemblies are contained in a containing vessel. An inert gas (N2, Ar or He) is filled in the containing vessel through an inert gas supply channel. The temperature of the inert gas is raised by the remaining after heat of the reactor core-constitutional elements. The inert gas is circulated and heated through a preheating circuit by driving a recycling gas blower and returned to the containing vessel. If the inert gas is heated to a predetermined temperature, metal sodium deposited on the surface of the materials of the reactor core-constitutional elements is evaporated. Next, a vacuum pump unit of a vacuum exhaustion channel is driven to suck an inert gas entraining sodium vapor in the containing vessel, and the sodium vapor is cooled, condensed thereby separated in a sodium separator. Then, the inert gas at a low temperature is introduced to a vacuum exhaustion channel to remove and discharge remained sodium vapor by a sodium trap. (I.N.)

  19. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  20. Feasibility study on ultrasonic flow meter for sodium-cooled FBR

    In the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems', 12-chromium steel which has the magnetic property is adopted to the primary and secondary system piping of sodium-cooled reactor. Hence the studies on ultrasonic flow-meter for primary and secondary cooling system have been conducted. In this study, elementary tests in the air and in water were performed in order to select ultrasonic transducer and couplant which suit high temperature and exchange by remote control operation and to verify the ultrasonic propagation characteristic of the transducer and couplant (mainly sensitivity) and the measurement performance in large-scale pipe (mainly influence of attenuation of ultrasonic). Furthermore, for conducing to the setup of set points like 'primary flow rate low' reactor trip signal detected by the ultrasonic flow-meter, the ultrasonic transducers for normal temperature were attached at the downstream in the elbow of the test facility which has been used for the water fluid and oscillation test for large diameter piping and fluctuation data from the ultrasonic transducers were acquired. The results are as follows: (1) As a result of investigating high temperature couplant, gold was selected since it was the most stable on all the temperature range containing high temperature among solid metal couplants and force required for the stability of the characteristic is expected to be small. (2) Signal level of ultrasonic transducer with the surface pressure of 1.45x104 Pa (the force of 5 kN) exceeds 50% of that with the couplant for normal temperature. In the transient temperature test, signal level of ultrasonic transducer became larger with a temperature rise and the fall of the signal level considered to originate in the fall of force was not appear. (3) S/N ratio of the signal exceeded 23dB in the elementary test in water, and it was expected more than 18dB even if more attenuation of ultrasonic by the propagation longer distance in sodium was

  1. A preliminary design study of a pool-type FBR 'ARES' eliminating intermediate heat transport systems

    An innovative reactor concept 'ARES' (Advanced Reactor Eliminating Secondary system) is proposed to aim at reducing the construction cost of a liquid metal cooled fast breeder reactor (LMFBR). This concept is developed to show the ultimate cost down potential of LMFBR's at their commercial stage. The electrical output is 1500 MW, while the thermal output is 3900 MW. Main components of the primary cooling system are four electromagnetic pumps (EMP) and eight double-wall-tube steam generators (SG). Both of them are installed in a reactor vessel like pool type LMFBR's. An intermediate heat transport system which a previous LMFBR has it eliminated, main components of which are intermediate heat exchangers (IHX), secondary pumps and secondary piping. Further, a high reliable SG could decrease the occurrence of water leak accidents and reduce the related mitigation systems. In this study, structure concept, approach to embody a high reliable SG and accidents analyses are carried out. Flow path configuration is mainly discussed in investigation of the structure concept. In case of a water leak accident in a SG, the fault SG must be isolated to prevent a reaction production from flowing into the core. The measure to cut both inlet and outlet coolant flow paths by siphon-break mechanism is adopted to be consistent with the decay heat removal operation. The safety design approach of the double-wall-tube SG is investigated to limit the accident occurrence below 10-7 (1/ry). A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to prevent adhesion of the double-wall-tube effectively. The reliability of the tube-to-tube-sheet was evaluated as 10-10 (1/hr) for an inner tube and 10-9 (1/hr) for an outer tube with reference to the failure experience of previous SG's. The failure must be detected within 60 to 120 minutes. Finally, a seamless U tube type of double-wall-tube SG is adopted. Transient events due to

  2. Safety evaluations of FBR power plants and fuel cycle facilities. Results of the studies in 2004

    This study is dedicated to establish rational safety design concepts for the power plants and their related fuel cycle facilities in the feasibility study on commercialized fast reactor cycle systems. Major results of this study are as follows. The principles of safety design and evaluation for the sodium-cooled reactor were formulated as well as the way to achieve the equivalent safety level to LWRs, design requirements to eliminate necessity of the evacuation, and the containment performance requirement. As the result, the item for further investigation was clarified in the containment design. As for small scale sodium-cooled reactor with metallic fuel, state-of-the-arts of knowledge for the evaluation of CDA, the way for elimination of re-criticality and related design requirements for the reactor core were identified. A preliminary evaluation showed that the feasibility of molten fuel discharge capability of modified inner duct concept for sodium-cooled MOX fuel reactors. The three dimensional effect in the course of the transition phase of CDA was investigated and found that the cooling effect of the control rod guide tube is pronounced comparing with two dimensional case. A preliminary evaluation showed that the occurrence probability of PLOHS can be reduced by more than one order with help of some accident management measures such as steam supply to SGs, improvement of diversity for the air cooler dumper. Concerning the SG tube rupture for the LBE-cooled reactor, it was found out that certain amount of steam can enter into the core in case of one tube rupture at the bottom of SG and that steam jet breaker should be installed in order to avoid massive steam ingress. For a gas-cooled reactor, a scenario of core disruptive accidents is qualitatively grasped to construct a concept of design measures such as core catcher. A risk study on some typical events in the facilities of the super-critical direct extraction reprocessing and of the extraction chromatography

  3. Study on the FBR cycle introduction scenario. 2. A study on the role of nuclear energy under the diversity of energy supply-and-demand

    This report concerns it self with the results of an investigation about the possibility of future nuclear utilization in the part of FBR Cycle Introduction Scenario Study in the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System (the F/S)'. We have investigated about the problems that confront energy industries and electric power companies, the capacities of distributed generation, the coexistence method of a distributed generation and large-scale power supply generation, and the development status of a small-scale nuclear reactor from a wide viewpoint. Especially the spread of distributed generation causes the decrease of the electricity demand which the electric power companies supplies. Since introduction scale of a distributed power supply is also expected to increase in the future, it will give some influences to a future nuclear plan and a power supply plan. The hydrogen utilization with out greenhouse gas mission is expected to spread with distributed generation, such as a fuel cell and a micro-gas turbine. Therefore, we proposed the new business model that the hydrogen produced by using nuclear surplus electricity is consumed distributed generation, such as a fuel cell and a micro-gas turbine. We plan to evaluate quantitatively the best power supply composition based on this load stability business model, FBR introduction capacities, the load factor, and the amount of CO2 reduction. (author)

  4. Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

    The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

  5. Analysis of fluid-structure interaction mechanism of a Na-FBR core while the evacuation of a gas pocket

    The purpose of this study is to improve the knowledge about the core behavior of a sodium fast breeder reactor (Na-FBR) during vibrations through the fluid-structure interaction analysis. Namely, we investigate the flowering of the Phenix core during the SCRAM for negative reactivity (AURN) and the seismic behavior of the core of Astrid project. Three approaches are followed: experimental campaign, performing of analytical solution and development of numerical model. We create a flow regime map to identify the flow regimes in the fluid gap for very short times scales (as AURN) as well as longer time scales (as seismic oscillations). The most suitable equation system (Navier-Stokes, Euler or linearized Euler) is chosen to model the fluid flow in the numerical code. To our knowledge, for the first time, an analytical solution for free vibration and very narrow gaps is proposed. We designed two experimental apparatus (PISE-1a and PISE-2c) composed respectively by 1 and 19 hexagonal assemblies (two crowns) of Poly-methyl methacrylate (PMMA). Every PMMA assembly is fixed to a stainless steel twin-blades support allowing only orthogonal oscillations with respect to generating line of assembly. The twin-blades supports are designed to give the same range frequency of Phenix assembly in liquid sodium. The experimental equipment PISE-1a is used to determine the dynamic characteristics of PISE-2c assembly, to calibrate instrumentation and for validating our numerical model. Free vibration tests in air are performed to evaluate the dynamic characteristics of the body. Free vibration experiments in water allow to assess the added mass and added damping effect on the frequency. Even though the fluid flow during vibration should be completely bidimensional, the fluid flow is affected by a 3D effect - named 'jambage' - at the top and the basis of the assembly. This effect produces a lower frequency than the theoretical value. Tests are modeled with a bidimensional

  6. Status of fast breeder reactor development in India

    The energy scenario and economic conditions in India are presented. India needs considerable energy for its rapid industrialisation with the liberal economic policy. Nuclear energy with FBR is the only large scale energy resource other than coal, available in the country. The present economic constraints have delayed the construction of new NPPs. The performance of operating reactors has improved considerably during the year. Operating experience of FBTR has been detailed particularly the reactivity incident and its investigations. Updated design of 500 MWe PFBR is presented. Various R and D works in support of FBR in the engineering, metallurgy, chemistry, reprocessing, safety etc. are detailed. (author)

  7. FBR type reactor

    A reaction product removing device is disposed between a primary cover gas region and a secondary cover gas resin, and a secondary plate is disposed each at the both ends thereof in addition to an existent distraction plate. If a heat-transfer pipe should be ruptured, sodium and water are reacted to generate a hydrogen gas, and a pressure in the secondary system is increased. The secondary destruction plate is destroyed by the pressure elevation, and the primary system and the secondary system are in communication with each other. In this case, the reaction products in the secondary system pass through the removing device to be removed with solid materials, and comprise only gases. Accordingly, since the pressure difference is reduced, rupture of a primary container can be prevented. In view of the above, integrity of the primary container can be kept and contamination of the primary system can be prevented. (N.H.)

  8. FBR type reactor

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  9. FBR type reactor

    A lower plate of lower seal bellows secured to the lower portion of an intermediate heat exchanger and a lower bellows seat mounted on the side of an inner cylinder constitute a lower seal mechanism. In addition, the lower plate of upper seal bellows and an upper bellows seat constitute an upper seal mechanism. A depressurizing mechanism is disposed to the lower seal mechanism. The depressurizing mechanism comprises a lower small hole formed at the lower end of the lower seal bellows and an upper small hole formed on the side of an exit plenum of a heat transfer pipe of the intermediate heat exchanger. Primary sodium pressurized by an electromagnetic pump flows from the lower small hole to the upper small hole and returned to the suction side of the electromagnetic pump by way of the exit plenum of the electromagnetic pump by way of the exit plenum of the heat transfer pipe of the intermediate heat exchanger. With such a constitution, pressure difference exerted on the sealing mechanism is mitigated. Accordingly, necessary conditions for the strength of the seal bellows can be moderated, as well as the amount of coolants leaked from the sealed portion can be decreased. (I.N.)

  10. Development of the ERIX process for reprocessing spent FBR-MOX fuel. Electrolytic reduction and anion exchange separation for U and FPs containing solution

    In recent years, we have been investigating the development of the ERIX process for reprocessing spent FBR-MOX fuel. The ERIX process uses electrolytic reduction and ion exchange techniques to recover U, Pu, Np and the minor actinides from spent EBR-MOX fuel solution. In previous work, it was found that U(VI) can be effectively reduced to U(IV) using the flow type electrolytic cell and U(IV) can be completely separated from fission products by AR-01 anion exchanger packed column in nitric acid medium. In the present work, electrolytic reduction behaviors of U(VI) and FPs containing solutions were investigated and the effects of various fission products, hydrazine, nitric acid and nitrous acid were examined. Furthermore, separation behavior of U from typical fission products in 6 mol/dm3 nitric acid solution after the electrolytic reduction were investigated using the column packed with AR-01. (author)

  11. Effect of Ce and La additions in low temperature aluminization process by CVD-FBR on 12%Cr ferritic/martensitic steel and behaviour in steam oxidation

    Two different coatings based of iron aluminide on 12% Cr ferritic-martensitic steel have been developed by CVD-FBR technique, which is modified by the introduction of Ce and La as powder in the fluidized bed. These elements change the gaseous environment, which composition is predicted by a thermodynamic approximation. Partial pressures of all gaseous precursors are drastically modified; in consequence AlCl has the highest partial pressure in the system leading to an increment of the coating thickness. Coatings are composed by (Fe, Cr)2Al5 or (Fe, Cr)2Al5 and (Fe, Cr)Al3 intermetallic phases. On the other hand, steam oxidation test at 650 deg. C was performed in order to observe improvements in the HCM12A oxidation resistant

  12. A New Methodology for Web-Knowledge- Based System Using Systematic Thinking, KM Process and Data & Knowledge Engineering Technology:FBR-GAs-CBRC5.0- CART

    Patcharaporn Paokanta

    2013-10-01

    Full Text Available In Knowledge Management perspective, Organization Learning and the selection of Knowledge Management tool affects the Knowledge Management strategy planning. Among the various KM theorem such as Learning method, organization knowledge creation, Cognitive theory, Intangible assets and knowledge capital, Measuring knowledge theory etc., Systematic Thinking plays an important role in Knowledge Management activities especially, the creation of Knowledge Management strategy, KM process and Knowledge Management system. DKET is one of several approaches for implementing the Knowledge Management tools based on the KM strategies. They are not only implemented in forms of standalone system but the web-online system also. Generally, DKET namely Ensemble Learning is well known as the technique of using different training data sets or learning algorithms. Currently, a popular learning algorithm is Fuzzy-Based Reasoning (FBR which the concept of this theory is “each item is notmatched to a given cluster but it has a degree of belonging to a certain cluster”. According to these reasons, in this paper, a new methodology for Web-Knowledge-Based System by using SystematicThinking, Knowledge Process and DKET (FBR-GAs-C5.0-CART is proposed in terms of KM perspective. The algorithm performance comparisons of Fuzzy C-Means-CBR-GAs-C5.0-CART in several data sets are presented. The satisfied clustering results of Fuzzy-C Means-GAs-CBR-C5.0-CARTattain RMSE at 5.10 for the case that full data set, on the other hand the best result of using Fuzzy-C Means-CBR-C5.0-CART attain RMSE at 12.03 in the case that unrecoded variables and CBR-C5.0- CART without symptoms variables. In the future, the other KM theories and DKET will be applied to improve the performance of this system.

  13. Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

    A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

  14. Liquid metal cooling issues for fusion and fission

    Liquid metal application to nuclear power plants was initiated in a design of fast reactors with using sodium or lead bismuth eutectic, and developed into a sodium fast breeder reactor and lead bismuth fast reactor. In the development stage, Na and NaK were carefully compared and the former was chosen. In the nuclear fusion application, liquid metals of Li or LiPb will be used as a coolant and tritium breeder. A nuclear reactor requires two materials of moderator and coolant. Water or sodium satisfies double duty, leading to the oligopoly situation by LWR or Na-FBR. The success of these reactors depends on the selection of coolant material that works as a moderator. On an analogy of this history, fusion power plant should be integrated to employ a coolant that works as tritium breeder, such as Li or LiPb. Technology progresses in the system design are introduced, which will have synergy effect for fusion

  15. Liquid metal cooling issues for fusion and fission

    Horiike, H. [Graduate School of Engineering, Osaka University, 2-1 Yamada-oka, Suita City, Osaka 565-0871 (Japan)], E-mail: horiike@nucl.eng.osaka-u.ac.jp; Konishi, S. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji City, Kyoto 611-0011 (Japan); Kondo, H.; Yamaguchi, A. [Graduate School of Engineering, Osaka University, 2-1 Yamada-oka, Suita City, Osaka 565-0871 (Japan)

    2008-12-15

    Liquid metal application to nuclear power plants was initiated in a design of fast reactors with using sodium or lead bismuth eutectic, and developed into a sodium fast breeder reactor and lead bismuth fast reactor. In the development stage, Na and NaK were carefully compared and the former was chosen. In the nuclear fusion application, liquid metals of Li or LiPb will be used as a coolant and tritium breeder. A nuclear reactor requires two materials of moderator and coolant. Water or sodium satisfies double duty, leading to the oligopoly situation by LWR or Na-FBR. The success of these reactors depends on the selection of coolant material that works as a moderator. On an analogy of this history, fusion power plant should be integrated to employ a coolant that works as tritium breeder, such as Li or LiPb. Technology progresses in the system design are introduced, which will have synergy effect for fusion.

  16. Behavior of an heterogeneous annular FBR core during an unprotected loss of flow accident: Analysis of the primary phase with SAS-SFR

    In the framework of a substantial improvement on FBR core safety connected to the development of a new Gen IV reactor type, heterogeneous core with innovative features are being carefully analyzed in France since 2009. At EDF R and D, the main goal is to understand whether a strong reduction of the Na-void worth - possibly attempting a negative value - allows a significant improvement of the core behavior during an unprotected loss of flow accident. Also, the physical behavior of such a core is of interest, before and beyond the (possible) onset of Na boiling. Hence, a cutting-edge heterogeneous design, featuring an annular shape, a Na-plena with a B4C plate and a stepwise modulation of fissile core heights, was developed at EDF by means of the SDDS methodology, with a total Na-void worth of -1 $. The behavior of such a core during the primary phase of a severe accident, initiated by an unprotected loss of flow, is analyzed by means of the SAS-SFR code. This study is carried-out at KIT and EDF, in the framework of a scientific collaboration on innovative FBR severe accident analyses. The results show that the reduction of the Na-void worth is very effective, but is not sufficient alone to avoid Na-boiling and, hence, to prevent the core from entering into the primary phase of a severe accident. Nevertheless, the grace time up to boiling onset is greatly enhanced in comparison to a more traditional homogeneous core design, and only an extremely low fraction of the fuel (<0.1%) enters into melting at the end of this phase. A sensitivity analysis shows that, due to the inherent neutronic characteristics of such a core, the gagging scheme plays a major role on the core behavior: indeed, an improved 4-zones gagging scheme, associated with an enhanced control rod drive line expansion feed-back effect, finally prevents the core from entering into sodium boiling. This major conclusion highlights both the progress already accomplished and the need for more detailed future

  17. Basic characteristics of bis(2-ethylhexyl)phosphate-impregnated adsorbent used for separation of minor actinides from FBR-spent fuel

    FBR-spent nuclear fuel includes a great deal of minor actinides (MA: Am and Cm), which become febrile. Radioactive wastes including MA require a large area of ground for dumping and result in high cost. In Fast Reactor Cycle System Technology Development Project (FaCT) in Japan, we have been investigating extraction chromatography for separation of long-lived MA and specific fission products (FP) from high-level liquid wastes (HLLW). This method is expected to allow us to reduce an organic solvent use and to realize compact equipment. In this work, we have studied the static and dynamic adsorption behavior of representative FP contained in HLLW, Mo(VI), Zr(IV), Nd(III) and EU(III), on a bis(2-ethylhexyl)phosphate (HDEHP)-impregnated adsorbent. Such fundamental data should facilitate the efficient design of efficient MA recovery processes. Column adsorption experiments with the HDEHP-impregnated adsorbent have revealed that an increase in a flow rate results in a short breakthrough time and reduces the adsorption capacity of the column for all the elements tested. These results strongly suggest that a lower flow rate is preferable to enhance the adsorption capacity of the adsorbent. (author)

  18. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about ±0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  19. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  20. X-ray excited optical luminescence, photoluminescence, photostimulated luminescence and x-ray photoemission spectroscopy studies on BaFBr:Eu

    Subramanian, N; Govinda-Rajan, K; Mohammad-Yousuf; Santanu-Bera; Narasimhan, S V

    1997-01-01

    The results of x-ray excited optical luminescence (XEOL), photoluminescence (PL), photostimulated luminescence (PSL) and x-ray photoemission spectroscopy (XPS) studies on the x-ray storage phosphor BaFBr:Eu are presented in this paper. Analyses of XEOL, PL and PSL spectra reveal features corresponding to the transitions from 4f sup 6 td sup 1 to 4f sup 7 configurations in different site symmetries of Eu sup 2 sup +. Increasing x-ray dose is seen to lead to a red shift in the maximum of the PL excitation spectrum for the 391 nm emission. The XEOL and XPS spectra do not show any signature of Eu sup 3 sup + in the samples studied by us, directly raising doubts about the model of Takahashi et al in which Eu sup 2 sup + is expected to ionize to Eu sup 3 sup + upon x-ray irradiation and remain stable until photostimulation. XEOL and PSL experiments with simultaneous x-ray irradiation and He - Ne laser excitation as well as those with sequential x-ray irradiation and laser stimulation bring out the competition betwe...

  1. Numerical investigation on the enhancement capability of annular chimney towards natural convective heat transfer in the interior zone of scaled down FBR core catcher

    Full text of publication follows: A numerical study has been carried out to determine the influence of annular cylindrical chimney on buoyancy-induced flow in the dished end cavity of scaled down Fast Breeder Reactor. Results are presented for (i) cylindrical chimney configuration and (ii) annular chimney configuration occupying the center of the circular plate. Two dimensional laminar simulations are obtained by solving the fully elliptical governing equations of flow and energy. The fluid is Newtonian and incompressible and satisfies the Boussinesq approximation. Results for the upward facing isothermal circular plate with chimney configurations in confined enclosure are analyzed. The velocity fields and isotherms are studied extensively to assess the impact of both geometries on the flow structure, dynamics and overall heat transfer characteristics in the cavity, towards enhancement of natural convective heat transfer. The predicted results for the cylindrical chimney are compared with known experimental results. The results are of interest to post accident heat removal in fast breeder reactors (FBR). (authors)

  2. Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

    The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

  3. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  4. Fast reactor technology development in china status and prospects

    China has decided to speed-up the nuclear power development. It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively. The basic strategy of PWR-FBR matched development with Fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted. Another strategy also decided is that the partitioning and transmutation of MA will be realized using fast burner and ADS. The fast reactor engineering development will be divided into three steps: China Experimental Fast Reactor (CEFR 65 MWt/20 MWe), 3China Prototype/Demonstration Fast Reactor (CPFR/CDFR ≥1 500 MWt/600 MWe) and China Demonstration Fast Breeder Reactor (CDFBR 1 000-1 500 MWe). The CEFR is under installation and pre-operation testing with it's first criticality planned in 2009. The design study of CPFR is just started in 2006. Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR. (authors)

  5. Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

    Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

    1998-01-01

    A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

  6. Utilization of eddy current flow meter for sodium flow measurement in FBRs

    Highlights: • Design and development of eddy current flow meters for FBRs have been carried out. • The developed flow meter was tested and calibrated in sodium. • The developed flow meter was used in FBTR reactor, India and in Phenix reactor, France. • The performance of the eddy current flow meter is found to be excellent. -- Abstract: Liquid metal cooled Fast Breeder Reactors form the second stage of Indian nuclear programme. Prototype Fast Breeder Reactor (PFBR) is 500MWe pool type reactor, which is under construction at Kalpakkam, India. Measurement of sodium flow in the reactor system is essential from both operational and safety point of view. Primary sodium pump discharge, secondary sodium circuit, safety grade decay heat removal circuit and auxiliary sodium circuits are the critical locations in an FBR, where sodium flow measurement is of importance. Apart from this, sodium flow measurement in subassemblies in the core facilitates the detection of flow blockage in the subassemblies. Permanent magnet flow meter is widely used for pipe flow measurement in sodium systems, however, it cannot be considered for applications like subassembly and pump discharge flow measurements. From the safety point of view, flow through the core should be ensured under all operating conditions in a FBR. Development of suitable flow sensors meeting the dimensional constraints and having accessibility for maintenance is a challenge. An eddy current flow meter was designed and developed for sodium flow measurement in PFBR and the developed probe was tested for its performance in a sodium test rig. The developed ECFM was also used for core flow measurements in Fast Breeder Test Reactor (FBTR), India and in Phenix reactor, France, during its end of life test before decommissioning. This paper discusses the activities related to the development, calibration, testing of ECFM probe for PFBR and results of its utilization for flow measurement in FBTR and in Phenix reactor, France

  7. Fabrication and evaluation of oxide dispersion strengthened (ODS) ferritic mother tubes for FBR fuel cladding tubes. Evaluation of the mechanical properties and recrystallization

    Mechanical properties and recrystallization characteristics of oxide dispersion strengthened ferritic mother tubes for FBR fuel cladding tubes were evaluated by micro-crystal direction analysis by electron back scattering pattern (EBSP), oxide particle dispersion parameter analysis and the mechanical property tests. The main results were as follows: (1) Microstructure. Although the ferritic steels were equi-axed grains in the transverse direction of tube, the aspect ratio increased with Ti and Y2O3 contents of the steel. The 20% cold worked tubes were high densities of dislocation. The martensitic steels were equi-axed grains both in transverse and longitudinal directions. M91 steel was a ferritic structure and low dislocation density. M92 steel was a full martensitic. (2) Tensile properties in the circumferential direction. The uniform elongation of ferritic steel was lowest at 400degC. In the martensitic steels, M92 steel with high Ti content showed a high strength and good ductility. (3) Creep rupture strength. For the ferritic steels, the internal creep rupture strength at 700degC became higher as Ti and Y2O3 contents increased. The martensitic steels showed high creep rupture strength by controlling the addition of Ti and Y2O3. (4) Recrystallization analysis of ferritic steels by EBSP. By the sequence of cold rolling and heat treatment, recrystallization became difficult. For a stable direction, (hkl) tends to form a strong texture. (5) Evaluation for distribution parameter of oxide particles. The threshold stress, which was calculated by the combination of the volume fraction of dispersoids and image analysis, exhibited reasonable results. (author)

  8. Gas entrainment in sodium cooled FBR. Preliminary simulation for 1:1.8 scale water upper plenum experiment by AQUA-VOF

    Prevention of gas entrainment from the free surface of a sodium pool is one of the important problems to be solved during the feasibility study of commercial fast breeder reactor (FBR). To study the phenomenon of gas entrainment, a 1:1.8 scale experimental model with water as working fluid is under preparation at JNC. Preliminary information related to the flow pattern and vorticity structure can be help for the experiment preparations. Thus, several types of numerical calculations of flow in upper plenum are underway using following codes: Finite-Element Method (FEM) code SPIRAL. Commercial code FLUENT with unstructured mesh. Finite-Difference Method (FDM) code AQUA-VOF. This report describes the preliminary flow field simulations in a simplified configuration of the 1:1.8 scale experimental model by AQUA-VOF that is a well-outdated version of the AQUA code. This version does not have appropriate model of turbulence and is designed for structural grids in Cartesian and cylindrical coordinates. Since the upper plenum area contains hot and cold legs, the boundary-fit meshing systems are impossible. However, AQUA-VOF version can predict behavior of the free surface. The results of the preliminary simulation indicated that the influence of the free surface on the flow field is not much large under the flow rate condition that would be set in the experiment, rather than boundary conditions. Although the strict reproduction of every vorticity structure is very difficult in such a simulation, it may be possible to predict the existence of vortices that might cause gas entrainment. (author)

  9. Obtención y caracterización de recubrimientos de aluminio con la incorporación de silicio sobre aceros ferriticos martensiticos (HCM-12A) mediante la tecnología CVD-FBR

    Francisco J Bolívar; Laura Sánchez; María P Hierro; Pérez, Francisco J.

    2013-01-01

    Los recubrimientos protectores son frecuentemente aplicados para incrementar la resistencia a la corrosión y a la oxidación de aceros ferritico-martensiticos. En este estudio, los recubrimientos de aluminio modificado con Si fueron depositados mediante deposición Química en Fase Vapor en Lecho Fluidizado a presión atmosférica (CVD-FBR) sobre el acero ferritico-martensitico HCM-12A. Los parámetros iniciales del proceso fueron optimizados mediante simulaciones termodinámicas realizadas con el s...

  10. Plans of verification tests for the ACTOR code analyzing fission products behavior in primary heat transportation system of FBR

    The ACTOR is the code to analyze the transfer behavior of fission products (FPs) in the coolant sodium and cover gas, which are released from the fuel plenum or fuel pellets due to a cladding breach caused by temperature increase during an accident. Principal analysis models adopted in the ACTOR code; such as 'FP release model', 'Bubble transfer model in sodium', 'FP transfer model from bubble to sodium' and 'FP adsorption model', are constructed based on the experimental results relating to each phenomenon But, some of the analysis models are still desirable to be verified by experiments considering influence on the evaluation results. This study is planned to carry out the necessary verification tests for the ACTOR code concerning the following phenomena relation to the analysis models. (1) Cesium transfer behavior from inert gas bubble to sodium (2) Cover gas radiation behavior during temperature increase due to decay heat of FPs As for (1), transfer behavior of cesium to coolant in fast reactor is different from that of LWR, because cesium is alkaline metal same as sodium. So, cesium is to be an important fission product in severe accident of fast reactor plant. The analysis model of ACTOR, transfer behavior of cesium to coolant sodium is conducted based on the results of tests carried out by using iodine as fission product, so it is necessary to carry out the tests by using cesium for verification of ACTOR analysis model. As for (2), radiation heat transfer is not considered during cover gas temperature increase in the ACTOR analysis model, because argon applied as cover gas in fast reactor plant is monoatomic molecule. It is necessary to confirm the validity of this analysis model and to investigate a means to control cover gas temperature. This paper will report the test plans and test apparatus of verification tests for the ACTOR code mentioned above. The verification tests will be carried out by Hokkaido University