Core dynamics of HTR under ATWS and accident conditions

International Nuclear Information System (INIS)

Nabbi, R.

1988-05-01

The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP) [de

Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

International Nuclear Information System (INIS)

Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun

2014-01-01

The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available

Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

Energy Technology Data Exchange (ETDEWEB)

Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)

2014-10-15

The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.

Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

International Nuclear Information System (INIS)

Beller, D.E.; Sailor, W.C.; Venneri, F.

1998-01-01

A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO 2 -fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO 2 fuel cycle

ATWS analysis for total loss of feedwater sequence in UCN 3 and 4

International Nuclear Information System (INIS)

Park, S. H.; Song, Y. M.; Kim, D. H.; Kim, S. D.; Park, S. Y.

1999-01-01

ATWS is a trip-failed severe accident initiated from the transients like a turbine trip, a control bank withdrawal, and a loss of feedwater which are expected to occur comparatively often (one or two occurrences / year). In this study, an ATWS sequence in Ulchin 3 and 4 is analyzed and the effects of the important systems are studied for accident management purpose using a MIDAS/PK computer code. The MIDAS/PK code has been developed via coupling a point kinetics module with the MELCOR code. The code calculates a primary peak pressure of about 24MPa at 240 seconds for the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. Along with the basic ATWS analysis, several sensitivity runs are performed. From these, the turbines and the safety depressurization system (SDS) are judged to be important. The turbine trip resulting in a loss of offsite power and a RCP trip, degrades primary heat transfer to the secondary sides, and in turn, increases primary coolant temperature which reduces the reactor power due to the negative moderator temperature coefficient. Manual operation of SDS has an effect to lower the primary peak pressure considerably via supplementary depressurization in addition to the PORVs

A Los Alamos concept for accelerator transmutation of waste and energy production (ATW)

International Nuclear Information System (INIS)

1990-01-01

This document contains the diagrams presented at the ATW (Accelerator Transmutation of Waste and Energy Production) External Review, December 10-12, 1990, held at Los Alamos National Laboratory. Included are the charge to the committee and the presentations for the committee's review. Topics of the presentations included an overview of the concept, LINAC technology, near-term application -- high-level defense wastes (intense thermal neutron source, chemistry and materials), advanced application of the ATW concept -- fission energy without a high-level waste stream (overview, advanced technology, and advanced chemistry), and a summary of the research issues

Effect of reactor conditions on MSIV-ATWS power level

International Nuclear Information System (INIS)

Diamond, D.J.

1987-01-01

In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam that flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip [an anticipated transient without scram (ATWS) event], there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor conditions affect the power level during an MSIV-ATWS event. The time of interest is the 20- to 30-min period when it is assumed that the reactor is in a quasi equilibrium condition with the water level and pressure fixed, natural circulation conditions and no control rod movement or significant boron in the core. The initial conditions of interest are the time of the cycle and the operating state

Quantification of operator actions during ATWS following MSIV closure

International Nuclear Information System (INIS)

Luckas, W.J. Jr.; O'Brien, J.N.; Perline, R.K.; Spettell, C.M.

1986-01-01

Brookhaven National Laboratory (BNL) assisted the Accident Sequence Evaluation Program (ASEP) by performing a Human Reliability Analysis (HRA) of the operations crew tasks during the Anticipated Transient Without Scram (ATWS) accident sequence with Main Steam Isolation Valve (MSIV) closure at the Peach Bottom Atomic Power Station, Unit 2. A detailed task analysis was performed based on consideration of staffing, team interaction, and control room layout at Peach Bottom. ATWS scenarios developed by Oak Ridge National Laboratory (ORNL) and Idaho National Engineering Laboratory (INEL) were reviewed. Discussions were held with thermal-hydrodynamic/core neutronics engineers at BNL to determine the success criterion for tasks. Five major operator tasks were identified. After reviewing a computerized data base of human error probabilities (HEPs) from 19 probabilistic risk assessments (PRAs) for tasks similar to those above to establish the historic range of HEPs for such errors, consensus opinion and structured expert judgment was used to quantify each of these tasks at each branch point in the event tree within that range

The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis

International Nuclear Information System (INIS)

Naumov, V.S.; Bychkov, A.V.

1995-01-01

The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF 2 -PuF 3,(4) - MAF n ): - continious removal of radioactive gases, volatile impurities and 'noble fission products'; - portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally

The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis

International Nuclear Information System (INIS)

Naumov, V. S.; Bychkov, A. V.

1995-01-01

The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF2-PuF3,(4)-MAFn): -continious removal of radioactive gases, volatile impurities and 'noble fission products'; -portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally

Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

International Nuclear Information System (INIS)

Giachetti, R.T.

1989-09-01

This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Varacalle, D.J.; Giri, A.M.; Koizumi, Y.; Koske, J.E.

1983-01-01

An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAPS/MOD1 computer code showed good agreement with the experimental data

The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF{sub 2} basis

Energy Technology Data Exchange (ETDEWEB)

Naumov, V.S.; Bychkov, A.V. [Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

1995-10-01

The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF{sub 2}-PuF{sub 3,(4)} - MAF{sub n}): - continious removal of radioactive gases, volatile impurities and {open_quotes}noble fission products{close_quotes}; - portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally.

Effects of buffer thickness on ATW blanket performances

International Nuclear Information System (INIS)

Yang, Won Sik

2001-01-01

This paper presents the preliminary results of target and buffer design studies for a lead-bismuth eutectic (LBE) cooled accelerator transmutation of waste (ATW) system, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using an 840 MWt LBE cooled ATW design, the effects of buffer thickness on the blanket performances have been studied. Varying the buffer thickness for a given blanket configuration, system performances have been estimated by a series of calculations using MCNPX and REBUS-3 codes. The effects of source importance change are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. As the irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. The results show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable

Target/blanket conceptual design for the Los Alamos ATW concept

International Nuclear Information System (INIS)

Ames, K.; Cappiello, M.; Ireland, J.; Sapir, J.; Farnum, G.

1992-01-01

The Los Alamos Accelerator Transmutation of Waste (ATW) concept has many potential applications that include defense waste transmutation, defense material production (i.e., tritium and 238 Pu), and the transmutation of hazardous nuclear wastes from commercial nuclear reactors (fission products and actinides). A more advanced long-term Los Alamos effort is investigating the potential of an accelerator- driven system to produce fission energy with a minimal nuclear waste stream. All applications employ a high-energy (800- to 1600-MeV), high-current (25--250 mA) proton linear accelerator as the driver. In this report, we discuss only the target/blanket conceptual design for the commercial nuclear waste application. A conceptual design for the target/blanket of the Los Alamos ATW concept has been presented. The neutronics, mechanical design, and heat transfer have been investigated in some detail for the base-case design. Much more work needs to be done, but at this point it appears that the design is feasible and will approach the design goal of supporting two commercial power reactors with each target/blanket module

BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

International Nuclear Information System (INIS)

Sehgal, B.R.; Peng, C.M.; Maly, J.

1988-01-01

The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

ATW neutron spectrum measurements at LAMPF

Energy Technology Data Exchange (ETDEWEB)

Butler, G.W.; Littleton, P.E.; Morgan, G.L. [Los Alamos National Laboratory, NM (United States)] [and others

1995-10-01

Accelerator transmutation of waste (ATW) is a proposal to use a high flux of accelerator-produced thermalized neutrons to transmute both fission product and higher actinide commercial nuclear waste into stable or short-lived radioactive species in order to avoid long-term storage of nuclear waste. At LAMPF the authors recently performed experiments that were designed to measure the spectrum of neutrons produced per incident proton for full-scale proposed ATW targets of lead and lithium. The neutrons produced in such targets have a spectrum of energies that extends up to the energy of the incident proton beam, but the distribution peaks between 1 and 5 MeV. Transmutation reactions and fission of actinides are most efficient when the neutron energy is below a few eV, so the target must be surrounded by a non-absorbing material (blanket) to produce additional neutrons and reduce the energy of high energy neutrons without loss. The experiments with the lead target, 25 cm diameter by 40 cm long, were conducted with 800 MeV protons, while those with the lithium target, 25 cm diameter by 175 cm long, were conducted with 400 MeV protons. The blanket in both sets of experiments was a 60 cm diameter by 200 cm long annulus of lead that surrounded the target. Surrounding the blanket was a steel water tank with dimensions of 250 cm diameter by 300 cm long that simulated the transmutation region. A small sample pipe penetrated the length of the lead blanket and other sample pipes penetrated the length of the water tank at different radii from the beam axis so that the neutron spectra at different locations could be measured by foil activation. After irradiation the activated foil sets were extracted and counted with calibrated high resolution germanium gamma ray detectors at the Los Alamos nuclear chemistry counting facility.

Calculation of the time behavior of a PWR NPP during a loss of feedwater ATWS case

International Nuclear Information System (INIS)

Hoeld, A.

1988-01-01

Event tree analyses of plant internal accidents play an important role within the safety evaluations of nuclear power reactors. The consequences after normal and abnormal operational perturbations have to be studied with respect to the safety situation of the entire plant and the possibility of additional failures in the reactor scram system be taken into account. In the analysis of anticipated transients with or without reactor scram (non-ATWS or ATWS-cases), it can, according to their initiating events, be distinguished between three important categories, namely - loss of off-site and on-site power (LOOP), - turbine-trip without opening of the bypass station, - loss of main feed water (LOFW). The last case with the additional assumption of a failure in the control rod drive will be subject of this presentation, calculating the dynamic behavior of a PWR NPP (with an end of cycle core, EOC) after such a LOFW/ATWS accident by the transient code combination ALMOD-4/UTSG-2. A short characterization of this combination will be given before consequences of such an accident and the interactions of the different plant parameters are discussed in more detail on basis of the corresponding calculation

Proposal on the accelerator driven molten-salt reactor (ATW concept) benchmark calculations. (STAGE 1 - without an external neutron source)

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

The first stage of ATW neutronic benchmark (without an external source), based on the simple modelling of two component concept is presented. The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark is not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (author)

Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.

1993-01-01

Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

Risk evaluation of the alternate-3A modification to the ATWS prevention/mitigation system in a BWR-4, MARK-II power plant

International Nuclear Information System (INIS)

Papazoglou, I.A.; Bari, R.A.; Karol, R.; Shiu, K.

1983-01-01

The authors present a risk evaluation of the ATWS Alternate 3A modification proposed by NRC staff in NUREG-0460 to the ATWS prevention/mitigation system in a BWR nuclear power plant. The evaluation is done relative to three risk indices: the frequency of core damage, the expected early fatalities, and the expected latent fatalities. The ATWS prevention tree includes: the mechanical subsystem of the reactor protection system, the electrical subsystem of the reactor protection system, the recirculation pump trip and the Alternate Rod Insertion System. The mitigation tree includes: standby liquid control system, opening of the relief valves, reclosing the relief valves, failure of coolant injection, inadvertent actuation of the automatic depressurization system, inadvertent operation of high-pressure injection system and containment heat removal

The feasibility study I on the blanket fuel options for the ATW/HYPER

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L

2001-01-01

The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended.

The feasibility study I on the blanket fuel options for the ATW/HYPER

International Nuclear Information System (INIS)

Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L.

2001-01-01

The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended

Effects of buffer thickness on ATW blanket performance

International Nuclear Information System (INIS)

Yang, W. S.; Mercatali, L.; Taiwo, T. A.; Hill, R. N.

2001-01-01

This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy ( and lt; 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level

Effects of Buffer Thickness on ATW Blanket Performance

International Nuclear Information System (INIS)

Yang, W.S.; Mercatali, L.; Taiwo, T.A.; Hill, R.N.

2002-01-01

This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level. (authors)

Simulation of BWR stability following an ATWS with boron injection using TRAC-BF1 with one-dimensional kinetics

International Nuclear Information System (INIS)

Lider, S.; Maclan, R.; Baratta, A.J.; Mahaffy, J.; Robinson, G.E.

2004-01-01

The scenario following an ATWS is characterized by the necessity to reduce the power in the reactor as fast as possible. The only means to insert a significant amount of negative reactivity in a BWR during an ATWS are the natural reactor negative void coefficient, and the injection of highly enriched boron through the SLCS. The ATWS management strategy suggested by BWR owner's group contemplates an initial rapid decrease in power as a result of the recirculation pump trip. This is followed by lowering of vessel water level and the injection of borated water into the lower plenum. A recent paper of Dias, et al. reports that reducing core power and lowering water level causes a reduction in boron mixing efficiency and the net effect is a longer time to shut down and an increase in Suppression Pool (SP) temperature. In the present paper, a series of analyses are made to address this issue. The preliminary results for the water level positions at TAF, TAF+1.5 m (TAF+5') and TAF+3 m (TAF+10') support the similar findings of Dias, et al. (author)

Investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a BWR ATWS

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Layman, W.; Hentzen, R.D.; Gose, G.C.

1985-01-01

A best-estimate analysis was performed to evaluate the technique of intentionally reducing reactor coolant inventory in order to reduce power during a BWR ATWS event. The ATWS was initiated by closure of the main steam isolation valves. The analysis was performed with the RETRAN-02 computer code utilizing the one-dimensional kinetics option. The one-dimensional cross sections were developed using the SIMULATE-E and SIMTRAN-E computer codes. The MSIV closure transient provides some of the more severe conditions following a postulated failure to scram. In this transient, the only mechanism for removing energy from the vessel is through the safety relief valve system which results in a heating up of the suppression pool fluid. Consequently, the reactor power must be reduced so that the suppression pool temperature limits are not exceeded. Under the proposed emergency procedure guidelines for the ATWS event, the reactor vessel water level will be lowered to reduce system power. This analysis evaluated the dynamic response of the system as the water level was lowered to the top of active fuel evaluation. Correlating the system power and flow patterns to water level was of particular interest in the analysis. Under natural circulating conditions, the system flows, core power, and pressure responses are extremely tightly coupled and the analysis results proved to be very sensitive to the modeling of downcomer, upper plenum, and core regions

Safety assessment of the Indonesian multipurpose reactor RSG-GAS against ATWS and hypothetical accidents

International Nuclear Information System (INIS)

Hastowo, H.; Nabbi, R.; Prayoto; Ismuntoyo, R.P.H.

2004-01-01

Investigation on ATWS and hypothetical accidents for the Indonesian Multipurpose Reactor RSG-GAS have been undertaken by computer simulation technique. Two computer codes, namely RELAP5 and PARET-ANL, were used as the main tools. The RELAP5 was utilized to perform system analysis while the PARET-ANL code was used to perform the reactor core analysis in more detail. Two different models have been applied as a basis of the simulation: Typical Working Core model (IWC-model) consisting of four regions with different radial power factors; and the hot-channel model consisting of two regions with different radial power factors. Both RELAP5 ad PARET-ANL results showed that in the occurrence of ATWS, failure on fuel element or fuel plate was limited to the region with the most highest power factor. The results also indicated that no high pressure development occurs in that region, so that mechanical damage on the fuel element or other core components due to pressure shock did not happen.(author)

Generic implications of ATWS events at the Salem Nuclear Power Plant: generic implications. Vol. 1

International Nuclear Information System (INIS)

1983-04-01

This report is the first of two volumes. It documents the work of an interoffice, interdisciplinary NRC Task Force established to determine the generic implications of two anticipated transients without scram (ATWS) at the Salem Nuclear Power Plant, Unit 1 on February 22 and 25, 1983. A second report will document the NRC actions to be taken based on the work of the Task Force. The Task Force was established to address three questions: (1) Is there a need for prompt action for similar equipment in other facilities. (2) Are NRC and its licensees learning the sefety-management lessons, and, (3) How should the priority and content of the ATWS rule be adjusted. A number of short-term actions were taken through Bulletins and an Information Notice. Intermediate-term actions to address the generic issues will be addressed in the separate report and implemented through appropriate regulatory mechanisms

ATWS: a reappraisal, part II, evaluation of societal risks due to reactor protection systems failure. Vol. 3. Pwr risk analysis. Phase report

International Nuclear Information System (INIS)

Lellouche, G.S.

1976-08-01

This document is the third volume of part 2 in a series of studies which will examine the basis for the problem of Anticipated Transients Without Scram (ATWS). The purpose of part 2 is an evaluation of societal risks due to RPS failure based on more current data and methodology than used in WASH-1270. This volume examines and documents the potential contribution to societal risk due to ATWS in the PWR. Volumes 1 and 2 described a similar analysis for the BWR

Actinide and Xenon reactivity effects in ATW high flux systems

International Nuclear Information System (INIS)

Woosley, M.; Olson, K.; Henderson, D.L.

1995-01-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

Actinide and xenon reactivity effects in ATW high flux systems

International Nuclear Information System (INIS)

Woosley, M.; Olson, K.; Henderson, D. L.; Sailor, W. C.

1995-01-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

Actinide and Xenon reactivity effects in ATW high flux systems

Energy Technology Data Exchange (ETDEWEB)

Woosley, M. [Univ. of Virginia, Charlottesville, VA (United States); Olson, K.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)] [and others

1995-10-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides.

Transmutation and inventory analysis in an ATW molten salt system

Energy Technology Data Exchange (ETDEWEB)

Sisolak, J.E.; Truebenbach, M.T.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)

1995-10-01

As an extension of earlier work to determine the equilibrium state of an ATW molten salt, power producing, reactor/transmuter, the WAIT code provides a time dependent view of material inventories and reactor parameters. By considering several cases, the authors infer that devices of this type do not reach equilibrium for dozens of years, and that equilibrium design calculations are inapplicable over most of the reactor life. Fissile inventory and k{sub eff} both vary by factors of 1.5 or more between reactor startup and ultimate convergence to equilibrium.

Estimation of the development possibility of the ABC/ATW fuel cycle based on LiF-BeF2 fuel salt. Part 2

International Nuclear Information System (INIS)

Bychkov, A.V.; Naumov, V.S.

1994-01-01

The aim of the first chapter was generalization of data on solubility and equilibrium states of fission product and actinide fluorides in fluoride salt melts-solvents and fuel composition melts based on LiF-BeF 2 mixture which was proposed as fuel basis for ABC/ATW facility. The second chapter is devoted to description of processes proposed for the chemical-technological complex of the ABC/ATW facility and their physico-chemical peculiarities. The complex is responsible for the removal of fission products and actinides from irradiated fuel salt

Evaluation of very low frequencies of ATWS and PLOHS in a loop-type FBR plant by making use of inherently safe features

International Nuclear Information System (INIS)

Sakata, K.; Koyama, K.; Aoi, S.; Simonelli, R.B.; Wallace, I.T.

1987-01-01

Frequencies of ATWS (Anticipated Transient Without Scram) and PLOHS (Protected Loss of Heat Sink) for a large loop-type FBR plant were evaluated by applying PSA methodologies. The frequencies were found to be so low that ATWS and PLOHS could be excluded from candidates of the design basis events. Furthermore, the inherently safe features introduced to the system design were verified to be very effective for reduction of the Probability of CCF (Common Cause Failure), which deteriorates reliability of both the reactor shutdown and the decay heat removal systems. (orig.)

Study on Fabrication of Ni-5 at.%W Tapes for Coated Conductors from Cylinder Ingots

DEFF Research Database (Denmark)

Ma, L.; Suo, H. L.; Yue, Zhao

2015-01-01

Ni-5 at.%W (Ni5W) tapes with a strong cube texture were fabricated using the RABiTS technique and by starting from cylindrical shaped ingots. In contrast to a conventional cuboid-shaped ingot, a cylinder shaped ingot has no anisotropy along the axial direction and the resulting tape will therefore...

Polonium release from an ATW burner system with liquid lead-bismuth coolant

International Nuclear Information System (INIS)

Li, N.; Yefimov, E.; Pankratov, D.

1998-04-01

The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

Environmental impacts of radiological consequences during the anticipated transients without scram (ATWS) events in nuclear power reactors

International Nuclear Information System (INIS)

El-Kafas, A.A.

2011-01-01

Anticipated transients without scram (ATWS), is one of the (worst case) accidents could happen if the system that provides a highly reliable means of shutting down the reactor (scram system )fails to work during a reactor event (anticipated transient).It has two general characteristics: (1) Initiation by a transient anticipated to occur one or more times in the life of reactor and ,(2) Assumed to proceed without scram.The types of events considered are those used for designing the plant .The evaluation of the radiological consequences during the assessment of the nuclear events,especially ATWS in nuclear power reactors, is very essential for environmental studies and public safety. In this paper, the root cases for nuclear events and dose calculation are presented. Scenario of accident sequences together with radiological impacts is illustrated for loss of coolant accident (LOCA) for a typical pressurized water reactor nuclear power plant. Recommendations for mitigating or preventing the release of radiation and high radioactive materials to environment are presented.

Direct irradiation of long-lived fission products in an ATW system

Energy Technology Data Exchange (ETDEWEB)

Carter, T.F. [Univ. of Tennessee, Knoxville, TN (United States); Henderson, D. [Univ. of Wisconsin, Madison, WI (United States); Sailor, W.C. [Los Alamos National Laboratory, NM (United States)

1995-10-01

The feasibility of directly irradiating five long-lived fission products (LLFPs: {sup 79}Se, {sup 93}Zr, {sup 107}Pd, {sup 126}Sn, and {sup 135}Cs, each with a half-life greater than 10,000 years), by incorporating them into the target of an Accelerator Transmutation of Waste (ATW) system is discussed. The important parameters used to judge the feasibility of a direct irradiation system were the target`s neutron spallation yield (given in neutrons produced per incident proton), and the removal rate of the LLFP, with the baseline incineration rate set at two light water reactors (LWRs) worth of the LLFP waste per year. A target was constructed which consisted of a LLFP cylindrical {open_quotes}plug{close_quotes} inserted into the top (where the proton beam strikes) of a 30 cm radius, 100 cm length lead target. {sup 126}Sn and {sup 79}Se were each found to have high enough removal rates to support two LWR`s production of the LLFP per year of ATW operation. For the baseline plug geometry (5 cm radius, 30 cm length) containing {sup 126}Sn, 3.5 LWRs could be supported per year (at 75% beam availability). Furthermore, the addition of a {sup 126}Sn plus had a slightly positive effect on the target`s neutron yield. The neutron production was 36.83 {plus_minus}.0039 neutrons per proton with a pure lead target having a yield of 36.29 {plus_minus}.0038. It was also found that a plug composed of a tin-selenide compound (SnSe) had high enough removal rates to burn two or more reactor years of both LLFPs simultaneously.

ATWS analysis for Browns Ferry Nuclear Plant Unit 1

International Nuclear Information System (INIS)

Dallman, R.J.; Jouse, W.C.

1985-01-01

Analyses of postulated Anticipated Transients Without Scram (ATWS) were performed at the Idaho National Engineering Laboratory (INEL). The Browns Ferry Nuclear Plant Unit 1 (BFNP1) was selected as the subject of this work because of the cooperation of the Tennessee Valley Authority (TVA). The work is part of the Severe Accident Sequence Analysis (SASA) Program of the US Nuclear Regulatory Commission (NRC). A Main Steamline Isolation Valve (MSIV) closure served as the transient initiator for these analyses, which proceeded a complete failure to scram. Results from the analyses indicate that operator mitigative actions are required to prevent overpressurization of the primary containment. Uncertainties remain concerning the effectiveness of key mitigative actions. The effectiveness of level control as a power reduction procedure is limited. Power level resulting from level control only reduce the Pressure Suppression Pool (PSP) heatup rate from 6 to 4 0 F/min

ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

International Nuclear Information System (INIS)

Parzer, I.; Kljenak, I.

2005-01-01

The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

Evolution of microstructure, texture and topography during additional annealing of cube-textured Ni–5at.%W substrate for coated conductors

DEFF Research Database (Denmark)

Wulff, Anders Christian; Mishin, Oleg; Grivel, Jean-Claude

2012-01-01

Microstructure, texture and topography have been studied in a recrystallized Ni–5at.%W substrate before and after additional annealing at 1025C for 1 h. The initial recrystallized material contained a strong cube texture and a high fraction of low angle grain boundaries. R3 boundaries were also f...

Proposal on the accelerator driven molten-salt reactor (ATW-concept) benchmark calculation (stage-1 without an external neutron sources)

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark will be not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (Authors)

Development of cube textured Ni-5 at.%W alloy substrates for coated conductor application using a melting process

International Nuclear Information System (INIS)

Zhao Yue; Suo Hongli; Liu Min; Liu Danmin; Zhang Yingxiao; Zhou Meiling

2006-01-01

Biaxially textured Ni-5 at.%W substrates have been prepared by cold rolling, followed by three different annealing routes. In this paper, the processes of melting Ni and W metals, flat rolling, various annealing methods are described in detail. The Ni-5 at.%W tapes annealed under either high vacuum or flowing Ar (7% H 2 ) gas were characterized by X-ray pole figures, ODF, EBSD as well as AFM analysis. The texture analysis indicated that as fabricated tapes have a sharp cube texture formed after annealing at a wide temperature range of 800-1100 o C. The high quality of cube orientation on tapes was obtained after a two-step annealing (TSA), where the percentage of the cube texture component was as high as 93.5% within a misorientation angle smaller than 8 o from EBSD analysis. Furthermore, it was also observed that the number of twin boundaries in this tape decreased with respect to that of tapes annealed both in vacuum and one-step gas annealing. From AFM on 1 μm 2 areas, it was concluded that the roughness (RMS) on the tape surface reached 0.98 nm

Estimates of thermal fatigue due to beam interruptions for an ALMR-type ATW

International Nuclear Information System (INIS)

Dunn, F. E.; Wade, D. C.

1999-01-01

Thermal fatigue due to beam interruptions has been investigated in a sodium cooled ATW using the Advanced Liquid Metal mod B design as a basis for the subcritical source driven reactor. A k eff of 0.975 was used for the reactor. Temperature response in the primary coolant system was calculated, using the SASSYS- 1 code, for a drop in beam current from full power to zero in 1 microsecond.. Temperature differences were used to calculate thermal stresses. Fatigue curves from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code were used to determine the number of cycles various components should be designed for, based on these thermal stresses

Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Johnsen, G.W.; Lellouche, G.S.

1994-03-01

The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

Parametric study on thermal-hydraulic response following as ATWS event

International Nuclear Information System (INIS)

Suh, Jeong Kwan; Bang, Young Seok; Kim, Hho Jung

2000-01-01

A series of sensitivity calculations for the LOFT L9-3 experiment were performed using RELAP5/MOD3 code to assess parametric effects on thermal-hydraulic response in the event of Anticipated Transient Without Scram (ATWS). The base case calculation was made by the condition which gave a good agreement for the pressure of the reactor coolant system (RCS) with the experimental data. Four parameters of PORV/spray energy loss coefficient, steam generator nodalization and moderator density coefficient (MDC) were selected during the input preparation and investigated by calculating the total discharged energy through relief valves. The energy loss coefficient of the pressurizer spray valve has a significant effect on the behavior of the RCS pressure and the change of the MDC curve within 15 % at the negative region decreased the difference of the coolant temperature between the experiment and the calculation within a range of measurement uncertainty. The finer steam generator nodalization increased the primary to secondary heat transfer rate

Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Grush, W.H.; Koizumi, Y.; Woerth, S.C.

1983-01-01

An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data

A simplified analysis of uncertainty propagation in inherently controlled ATWS events

International Nuclear Information System (INIS)

Wade, D.C.

1987-01-01

The quasi static approach can be used to provide useful insight concerning the propagation of uncertainties in the inherent response to ATWS events. At issue is how uncertainties in the reactivity coefficients and in the thermal-hydraulics and materials properties propagate to yield uncertainties in the asymptotic temperatures attained upon inherent shutdown. The basic notion to be quantified is that many of the same physical phenomena contribute to both the reactivity increase of power reduction and the reactivity decrease of core temperature rise. Since these reactivities cancel by definition, a good deal of uncertainty cancellation must also occur of necessity. For example, if the Doppler coefficient is overpredicted, too large a positive reactivity insertion is predicted upon power reduction and collapse of the ΔT across the fuel pin. However, too large a negative reactivity is also predicted upon the compensating increase in the isothermal core average temperature - which includes the fuel Doppler effect

Effect of power oscillations on suppression pool heating during ATWS [Anticipated Transients Without Scram] conditions

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.

1990-01-01

Nine selected Anticipated Transients Without Scram (ATWS) have been simulated on the BNL Engineering Plant Analyzer (EPA), to determine how power and flow oscillations, similar to those that did or could have occurred at the LaSalle-2 boiling Water Reactor (BWR), could affect the rate of Pressure Suppression Pool heating. It has been determined that the pool can reach its temperature limit of 80 degree C in 4.3 min. after Turbine Trip without Bypass, if the feedwater pumps are not tripped. The pool will not reach its limit, if Boron is injected, even when oscillations are encountered. Simultaneous turbine and recirculation pump trips, introduced under stable conditions, can lead to instability. 2 refs., 17 figs., 9 tabs

Induced structural radioactivity inventory analysis of the base case aqueous ATW reactor concept

International Nuclear Information System (INIS)

Bezdecny, J.A.; Henderson, D.L.; Sailor, W.C.

1993-01-01

The purpose of the Los Alamos National Laboratory Accelerator Transmutation of Nuclear Waste (ATW) project is the substantial reduction in volume of this country's long-lived high-level radioactive waste in a safe and energy efficient manner. An evaluation of the Accelerator Transmutation of Nuclear Waste concept has four aspects; material balance, energy balance, performance and cost. An evaluation of the material balance compares the amount of long-lived high-level waste transmuted with the amount and type of waste created in the process. One component of the material balance is the activation of structural materials over the lifetime of the transmutation reactor. An activation analysis has been performed on four structure regions of the reaction vessel: the tungsten target; the lead target and annulus; the Zircalloy and aluminum tubing carrying the actinide slurry and; the stainless steel tank

Generic implications of ATWS events at the Salem Nuclear Power Plant. Licensee and staff actions

International Nuclear Information System (INIS)

1983-08-01

This report, Volume 2 of two volumes of NUREG-1000, describes the intermediate term actions to be taken by licensees and applicants of the US Nuclear Regulatory Commission (NRC), on the one hand, and by NRC staff, on the other, to address the generic issues raised by two anticipated transients without scram (ATWS) at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983. These actions came about as a result of the findings of NUREG-1000, Volume 1, and of reviews by the NRC Committee to Review Generic Requirements, the NRC Program Offices, and the Commission. The actions to be taken by licensees and applicants have been detailed in a letter pursuant to 10 CFR 50.54(f)

Fabrication of textured Ni–9.3at.%W substrate by electropulsing intermediate annealing method

International Nuclear Information System (INIS)

Liu, Jianan; Liu, Wei; Tang, Guoyi; Zhu, Rufei

2014-01-01

Highlights: •It’s the first time that EIA is used on Ni9 W substrate production. •Compared with CIA, EIA trends to sharpen the rolling texture. •Improved cube recrystallization texture is obtained by EIA. •EIA provides a highly efficient approach for Ni9 W substrate manufacture. -- Abstract: Sharp cube texture is difficult to obtain in high W content Ni–W alloy substrates used for coated conductors. In this paper, a new method called electropulsing intermediate annealing (EIA) is adopted to optimize the rolling and recrystallization texture of Ni–9.3 at.%W substrate. It is found that, compared with conventional intermediate annealing (CIA) at the same temperature, EIA trends to increase the Copper, S and Brass components, suppress the Goss component in rolling texture. Higher cube recrystallization texture is obtained at relatively low temperature by EIA in a shorter time. The effect of EIA on texture is attributed to the enhancement of recovery process resulting from the athermal effects

Performance comparison of liquid metal and gas cooled ATW system point designs

International Nuclear Information System (INIS)

Yang, W.S.; Taiwo, T.A.; Hill, R.N.; Khalil, H.S.; Wade, D.C.

2001-01-01

As part of the Advanced Accelerator Application (AAA) program in the U.S., preliminary design studies have been performed at Argonne National Laboratory (ANL) and Los Alamos National Laboratory (LANL) to define and compare candidate Accelerator Transmutation of Waste (ATW) systems. The studies at ANL have focused primarily on the transmutation blanket component of the overall system. Lead-bismuth eutectic (LBE), sodium, and gas cooled systems are among the blanket technology options currently under consideration. This paper summarizes the results from neutronics trade studies performed at ANL. Core designs have been developed for LBE and sodium cooled 840 MWt fast spectrum accelerator driven systems employing re-cycle. Additionally, neutronics analyses have been performed for a helium-cooled 600 MWt hybrid thermal and fast spectrum system proposed by General Atomics (GA), which is operated in the critical mode for three cycles and in a subcritical accelerator driven mode for a subsequent single cycle. For these three point designs, isotopic inventories, consumption rates, and annual burnup rates are compared. The mass flows and the ultimate loss of transuranic (TRU) isotopes to the waste stream per unit of heat generated during transmutation are also compared on a consistent basis. (author)

Joyo ATWS test analysis by Mimir-N2

International Nuclear Information System (INIS)

Yoshida, Akihiro

2001-03-01

The study on the passive safety test by using the Experimental Fast Reactor Joyo was performed to demonstrate the inherent safety of fast breeder reactors. An analysis code: Mimir-N2, which has been developed to analyze Joyo plant kinetics, was selected as a standard code for this study. In order to increase the reliability of the calculation, Mimir-N2 code was adjusted based on the data obtained through several plant characteristics tests carried out in Joyo. Throughout an operational data obtained in Joyo, it is supposed that the burn-up dependency observed on the power reactivity coefficient might be coming from the reactivity shift caused by a depression of a thermal expansion of fuel pellet. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, the burn-up dependency mentioned above was estimated and introduced to Mimir-N2. As a result, calculated core and plant dynamics during the step reactivity response test, such as the response of the power range neutron monitor and the coolant temperature at the core inlet/outlet, corresponded with the measured value. Especially, it was confirmed that Mimir-N2 can simulate the perturbation caused by the thermal expansion of the core support plate. In addition, Mimir-N2 was modified to be enable to take into account for the core bowing reactivity, which is calculated by the core bowing reactivity analysis system developed for Joyo. The preliminary analysis of the plant dynamics during the ATWS events in MK-III core were carried out by using modified Mimir-N2. As a result, it was confirmed that the core bowing reactivity should not be neglected because it sometimes shows positive feedback characteristics. (author)

NRC Fact-Finding Task Force report on the ATWS event at Salem Nuclear Generating Station, Unit 1, on February 25, 1983

International Nuclear Information System (INIS)

1983-03-01

An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to an anticipated transient without scram (ATWS) event at the Public Service Electric and Gas Company's Salem Generating Station, Unit 1 on February 25, 1983. The charter of the Task Force was to determine the factual information pertinent to management and administrative controls which should have ensured proper operation of the reactor trip breakers in the solid state protection system. This report documents the findings of the Task Force along with its conclusions

Containment venting as a mitigation technique for BWR MARK I plant ATWS

International Nuclear Information System (INIS)

Harrington, R.M.

1987-01-01

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure

Processing flowsheet for the accelerator transmutation of waste (ATW) program

International Nuclear Information System (INIS)

Dewey, H.; Walker, R.; Yarbro, S.

1992-01-01

At Los Alamos, an innovative approach to transmuting long-lived radioactive waste is under investigation. The concept is to use a linear proton accelerator coupled to a solid target to produce an intense neutron flux. The intense stream of neutrons can then be used to fission or transmute long-lived radionuclides to either stable or shorter-lived isotopes. For the program to be successful, robust chemical separations with high efficiencies (>10 5 ) are required. The actual mission, either defense or commercial, will determine what suite of unit operations will be needed. If the mission is to process commercial spent fuel, there are several options available for feed preparation and blanket processing. The baseline option would be an improved PUREX system with the main alternative being the current ATW actinide blanket processing flowsheet. 99 Tc and 129 I are more likely to reach the biosphere than the actinides. Many models have been developed for predicting how the radionuclides will behave in a repository over long time periods. The general conclusion is that the actinides will be sorbed by the soil. Therefore, over a long time period, e.g., a million years their hazard will be lessened because of radioactive decay and dispersion. However, some of the long-lived fission products are not sorbed and could potentially reach the environment over a few thousand year period. Hence, they could present a significant safety hazard. Because of limited resources, most of the priority has been focused on the actinide and technetium blanket assemblies

Analysis of a high pressure ATWS [anticipated transient without scram] with very low make-up flow

International Nuclear Information System (INIS)

Wagner, K.C.

1988-10-01

A series of calculations were performed to analyze the response of General Electric Company's (GE) advanced boiling water reactor (ABWR) during an anticipated transient without scram (ATWS). This work investigated the early plant response with an assumed failure or manual inhibit of the high pressure core flooder (HPCF). Consequently, the reactor core isolation cooling (RCIC) and control rod drive (CRD) systems are the only sources of high pressure injection available to maintain core cooling. Steam leaving the reactor pressure vessel was diverted to the pressure suppression pool (PSP) via the steam line and the safety relief valves. The combination of an unscrammed core and the CRD and RCIC injection sources make this a particularly challenging transient. System energy balance calculations were performed to predict the core power and PSP heat-up rate. The amount of vessel vapor superheat and the PSP temperature were found to significantly affect the resultant core power. Consequently, detailed thermal-hydraulic calculations were performed to simulate the system response during the postulated transient. 15 refs., 15 figs., 4 tabs

Introducing a European Partnership. First issue of 'European Nuclear Features'. A joint publication of atw, Nuclear Espana, Revue Generale Nucleare (2004)

International Nuclear Information System (INIS)

2004-01-01

'European Nuclear Features' is a joint publication of the three specialized technical journals, Nuclear Espana (Spain), Revue Generale Nucleaire (France), and atw - International Journal for Nuclear Power (Germany), planned for six issues annually. ENF is to further greatly the international European exchange of information and news about energy and nuclear power. News items, comments, and scientific and technical contributions will cover important aspects of the field. The first issue of ENF contains contributions about these topics, among others: - European Nuclear Society and Foratom: Strengthening the Nuclear Network. - Report: EPR - the European Pressurized Water Reactor. - Finland: Starting Construction of the Fifth Nuclear Power Plant. - Czech Republic: Nuclear Power Report for 2003/2004. - The Decommissioning Project of the Bohunice-1 and -2 Units. - FRM-II: TUM Research Neutron Source Generates Its First Neutrons. (orig.)

RADDA - Comparison of results of three ATWS/ATWC scenarios simulated with the help of POLCA-T and S3K/RELAP5

International Nuclear Information System (INIS)

Peltonen, J.

2008-03-01

The effects of ATWS and ATWC-events with control rods failing to enter the core has been evaluated in this project. To understand the uncertainties in using modern 3D-calculation methods two different codes were used in the project. The outputs from the two code packages were compared. Within the project the used code were first evaluated against a real event, pancake core at Forsmark 3. The results give important knowledge of the core responses for such events and on how to use different code to perform such calculations. The NKS report is only one minor part of the total project. The project was sponsored by TVO, Forsmark, OKG, Ringhals, SKI besides the NKS-funding. The results could be used for PSA-studies and for deterministically safety analysis. (au)

Reliability and availability considerations in the RF systems of ATW-class accelerators

Energy Technology Data Exchange (ETDEWEB)

Tallerico, P.J.; Lynch, M.T.; Lawrence, G. [Los Alamos National Laboratory, NM (United States)

1995-10-01

In an RF-driven, ion accelerator for waste transmutation or nuclear material production, the overall availability is perhaps the most important specification. The synchronism requirements in an ion accelerator, as contrasted to an electron accelerator, cause a failure of an RF source to have a greater consequence. These large machines also are major capital investments, so the availability determines the return on this capital. RF system design methods to insure a high availability without paying a serious cost penalty are the subject of this paper. The overall availability goal in the present designs is 75% for the entire ATW complex, and from 25 to 35% of the unavailability is allocated to the RF system, since it is one of the most complicated subsystems in the complex. The allowed down time for the RF system (including the linac and all other systems) is then only 7 to 9% of the operating time per year, or as little as 613 hours per year, for continuous operation. Since large accelerators consume large amounts of electrical power, excellent efficiency is also required with the excellent availability. The availability also influences the sizes of the RF components; smaller components may fail and yet the accelerator may still meet all specifications. Larger components are also attractive, since the cost of an RF system usually increases as the square root of the number of RF systems utilized. In some cases, there is a reliability penalty that accompanies the cost savings from using larger components. The authors discuss these factors, and present an availability model that allows one to examine these trade offs, and make rational choices in the RF and accelerator system designs.

Nuclear energy and politics in Russian ATWS conditions

International Nuclear Information System (INIS)

Gagarinski, A.

1998-01-01

Relations between politics and nuclear power in the countries of sustainable development has been many times discussed during the short history of nuclear energy, and regularly arising new events, even very important (in Sweden, USA, etc.), just add to the formed understanding of the problem. Russia for 10 years lives in conditions of a transition period, which seems similar to ATWS-type accidents at nuclear power plants. In these conditions the effect of politics on nuclear power and vice versa are seen very clearly, and, more important, change swiftly, which may present interest for the countries with smoother public processes. The role of political processes in nuclear power is obvious and may be reduced to three main factors: change of political system and transition to market economy have placed nuclear power, though still within state sector, in an absolutely new economic condition, which determine its today's situation as 'Survival'; new possibilities of political influence and opposition to nuclear power (mainly struggle against construction of new nuclear fuel cycle objects) on a levels of authority (local, regional, federal); impact of the USSR collapse on the situation in Russian nuclear power was due sooner to temporary weakening of control and regulatory structures, than to the fact, that some fuel cycle elements have found themselves abroad (the factor of uranium resources' loss is unimportant at present). Nuclear safety was chosen to be the subject of Moscow 1996 Summit, initiated with the purpose of Russia coming closer to G7. The Summit has confirmed the thesis on the possibility of nuclear power o play an important role in the world energy demand in accordance with sustainable development goals. successful activities of Russia-USA Commission for economic and technological cooperation, known as 'Gore-Chernomyrdin' Commission, is to a large extent determined by positive nuclear decisions. Eastern direction of Russian nuclear export (Iran, China

Anticipated transients without scram for light water reactors. Appendices. Staff report

International Nuclear Information System (INIS)

1978-04-01

Information is presented concerning scram failure probability, rod drive failure data, ATWS rule and ATWS requirements, treatment of steam generator tube failures in ATWS evaluation, radiological consequences assessments, ATWS study to include parameter variations and equipment reliability in probabilistic accident analysis, PWR MTC for ATWS, safety valve flows, ATWS contribution to risk, fuel integrity, value-impact analysis, and analytical methods

Effect of Initial Surface Quality on Final Roughness and Texture of Annealed Ni-5at.%W Tapes Coated with a Gd2Zr2O7 Buffer Layer

DEFF Research Database (Denmark)

Wulff, Anders Christian; Yue, Zhao; Mishin, Oleg

2012-01-01

Surface roughness of Ni-5at.%W tapes in coldrolled and annealed conditions after subsequent deposition of a Gd2Zr2O7 buffer layer has been studied as a function of the polishing grade, taking grain boundary grooving into account. It is found that annealing decreases the initial mean surface...... roughness achieved by mechanical polishing of the cold-rolled material, except after very fine polishing. Furthermore, compared to the surface of the tape annealed after fine polishing, the mean roughness slightly increases after the deposition of the buffer layer. Grain boundary grooving was found...... to impose a lower limit for the mean surface roughness. In the annealed tapes, the fraction of orientations within 5◦ from the ideal cube orientation was observed to be very sensitive to the surface roughness before annealing....

Investigations of operational incidents without reactor scram (ATWS) and other selected safety devices

International Nuclear Information System (INIS)

Ullrich, W.; Frisch, W.

1976-09-01

The most important results may be summarized as follows: Analyses performed up to now show that the primary system is not directly endangered by any overpressure, as 1,1 times the design pressure will not be exceeded in light water reactors. In smaller areas of the reactor core, hazards may exist for several fuel rods. Here, additional tests are still required, especially during failures with a considerable lowering of the water level in the pressure vessel of BWRs or during failures with a very high steam development combined with pump failures in a PWR. Generally, computer models used are suitable to perform ATWS analyses. The confidence in the relatively recent PWR-models should be confirmed by comparison with other models and by reexaminations. Reliability studies of pressure relief systems and of those systems functioning in case of a scram, generally reveal that systems are of a high quality design. Deficiencies insofar as they have been recognized in time, have been eliminated during the licensing procedure. The determined nonavailability data for reactor scrams (RESA) are between 2 x 10 -6 and 5 x 10 -6 . Quantitive treatment of common mode failures is very difficult. First attempts for a solution have been made and results are given in chapters 8 and 9. More extensive studies should be performed in order to adequately quantify the common mode failures and in order to permit them to be handled as an integral part of reliability analyses. Results of analyses performed for BWRs and PWRs led to the conclusion that additional hardware measures on a large scale are not necessary now. Chapter 10, however, proposes possible improvements concerning the existing engineered safegurads for both the BWR and the PWR. These proposals should be discussed with the RSK and manufacturers and utilities as well, in order to achieve an optimum safety standard and to avoid a priori any adverse effect. (orig./HP) [de

Anticipated transients without scram for WWER reactors. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1999-12-01

Anticipated transients without scram (ATWS) are anticipated operational occurrences followed by the failure of one reactor scram function. Current international practice requires that the capability of pressurized water reactors (PWRs) to cope with ATWS be demonstrated following a systematic evaluation of plants' defence in depth. Countries operating PWRs require design consideration of ATWS events on a deterministic basis. The regulatory requirements may concern either specific mitigating systems or acceptable plant performance during these events. The prevailing international practice for performing transient analysis of ATWS for licensing is the best estimate approach. Available transient analyses of ATWS events indicate that WWER reactors, like PWRs, have the tendency to shut themselves down if the inherent nuclear feedback is sufficiently negative. Various control and limitation functions of the WWER plants also provide a degree of defence against ATWS. However, for most WWER plants, complete and systematic ATWS analyses have yet to be submitted for rigorous review by the regulatory authorities and preventive or mitigative measures have not been established. In addition, it has also been recognized that plant behaviour in case of ATWS also relies on certain system functions (use of pressurizer safety valves for liquid discharge, availability of steam dump valve to both the condenser (BRU-K) and the atmosphere (BRU-A) for secondary side pressure control, and others) which have been identified as safety issues and need to be qualified for accident conditions. In all countries operating WWERs, the need for ATWS investigations is recognized and reflected in the safety improvement programmes. ATWS analysis for WWERs is not required for the licensing process in Bulgaria, the Czech Republic (with the exception of the Temelin nuclear power plant) and Russia. Design consideration of ATWS is required if expert assessments of probabilistic safety assessment (PSA) results

CONTEMPT/LT-028 Browns Ferry studies of an anticipated transient without scram

International Nuclear Information System (INIS)

Holcomb, E.E.

1983-01-01

The Browns Ferry Nuclear Plant containment response during the first 30 min of an anticipated transient without scram (ATWS) is the subject of this paper. Three cases, each initiated by a main steam isolation valve closure, are presented: the ATWS is mitigated by operator actions in the spirit of the General Electric Emergency Procedure Guidelines; the ATWS is managed by the plant automatic control systems; and the ATWS proceeds as in first case except that the drywell coolers are unavailable. Success of the standby liquid control system is assumed in the last two transients

EP1000 anticipated transient without scram analyses

International Nuclear Information System (INIS)

Saiu, G.; Frogheri, M.; Schulz, T.L.

2001-01-01

The present paper summarizes the main results of the Anticipated Transient Without Scram (ATWS) analysis activity, performed for the European Passive Plant Program (EPP). The behavior of the EP1000 plant following an ATWS has been analyzed by means of the RELAP5/Mod3.2 code. An ATWS is defined as an Anticipated Transient accompanied by a common mode failure in the reactor protection system, such that the control rods do not scram as required to mitigate the consequences of the transient. According to the experience gained in PWR design, the limiting ATWS events, in a PWR, have been found to be the heatup transients caused by a reduction of heat removal capability by the secondary side of the plant. For this reason, the Loss of Normal Feedwater initiating event, to which the failure of the reactor scram is associated, has been analyzed. The purpose of the study is to verify the performance requirements set for the core feedback characteristics (that is to evaluate the effect of the low boron core neutron kinetic parameters), the overpressure protection system, and boration systems to cope with the EUR Acceptance Criteria for ATWS. Another purpose of this analysis was to support development of revised PSA success criteria that would reduce the contribution of ATWS to the large release frequency (LRF). The low boron core improved the basic EP1000 response to an ATWS event. In particular, the peak pressure was significantly lower than that which would result from a standard core configuration. The improved ATWS analysis results also permitted improved ATWS PSA success criteria. For example, the reduced peak pressure allows the use of other plant features to mitigate the event, including manual initiation of feed-bleed cooling in the event of PRHR HX failure. As a result, the core melt frequency and especially the LRF are significantly reduced. (author)

Abstracts of papers from the literature on anticipated transients without scram for light water reactors 1. 1975-1979

International Nuclear Information System (INIS)

Kinnersley, S.R.

1981-05-01

INIS ATOMINDEX abstracts relating to ATWS for light water reactors for the years 1975-1979 are presented under the subject headings of; general, licensing and standards, models and computer codes, frequency of occurrence of ATWS, transient calculations of results including probabilistic analysis, radiological consequences of ATWS, fuel behaviour, and studies of plant components. (U.K.)

An Estimation of Risk Impact of Anticipated Transients without Scram for a KSNP

Energy Technology Data Exchange (ETDEWEB)

Han, Seok Jung; Yang Joon Eon

2006-07-15

Anticipated transient without scram (ATWS) event is an accident sequence with large risk impact, while it is a beyond design basis accident (BDBA). We have estimated a risk due to an ATWS accident sequence for the KSNP in consideration of the recent accident analysis results. The SECY-83-293's model for the CE type plants has been used in a risk estimation of ATWS. A risk estimation due to an ATWS for the KSNP has been performed in consideration of the recent ATWS accident analysis results and plant information. We reviewed influence factors in the SECY-83-293's model, these factors have been re-estimated by using current information and PSA results for a KSNP. A risk due to an ATWS has been estimated as 3.6E-6/yr of CDF by using domestic aspect and recent KSNP information. A sensitivity study for the UET variation has been performed. As the results of the sensitivity analysis, the overall risk spectrum by the UET variation is bounded between 7.80E-7/yr to 8.00E-6/yr of CDF. As the result of the current study, the risk due to an ATWS accident sequence has been identified as a considerable impact on the entire risk of a KSNP, so the risk estimation of that plant should be upgraded by considering the recent information like the ATWS accident analysis results. Finally, we expect that this study can become a basis for the entire risk estimation of the referred plant.

A novel approach using powder metallurgy for strengthened RABiTS composite substrates for coated superconductors

International Nuclear Information System (INIS)

Suo Hongli; Zhao Yue; Liu Min; Ye Shuai; Zhu YongHua; He Dong; Ma Lingji; Ji Yuan; Zhou Meiling

2008-01-01

We report on the development of mechanically strengthened, highly textured Ni-5 at.%W/Ni-12 at.%W composite materials prepared by a powder metallurgical approach as promising weakly magnetic substrates for coated superconductors. The key configuration of this composite substrate consists of a thin, sharp cubic textured Ni-5 at.%W layer on a Ni-12 at.%W alloy core, thus providing a mechanical reinforcement while decreasing the saturation magnetization of the whole substrate. The composite substrates have a sharp cubic texture at the top Ni-5 at.%W outer layer and their yield strength reaches 272 MPa, exceeding that of the commercially used Ni5W substrates by a factor of 1.6. The saturation magnetization of the composite substrate Ni5W/Ni12W/Ni5W is substantially reduced when compared to that of pure Ni and Ni-5 at.%W substrates, respectively

MIDAS/PK code development using point kinetics model

International Nuclear Information System (INIS)

Song, Y. M.; Park, S. H.

1999-01-01

In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

Anticipated transient without SCRAM experiments at LOFT

International Nuclear Information System (INIS)

Grush, W.H.; Harvego, E.A.; Koizumi, Y.; Varacalle, D.J.

1983-01-01

This paper discusses the experimental results for two anticipated transients without scram (ATWS) experiments, and compares computer code predictions with the experimental data. Experiment L9-3 simulated an ATWS in a commercial pressurized water reactor (PWR) initiated by a complete loss of feedwater and Experiment L9-4 simulated a loss-of-offsite-power-initiated (loss of feedwater and trip of the primary coolant pumps) ATWS. The LOFT facility is uniquely suited for ATWS experiments because it is a volumetrically scaled (1/44) experimental PWR designed to simulate the major components and system responses of larger commercial PWRs during both hypothesized loss-of-coolant accidents and anticipated transients. In both of the examined experiments, the primary system transient behavior was dominated by the interactions between the steam generator primary-to-secondary heat removal, the reactor kinetics, and the relief valve actuation. It is demonstrated that the discussed ATWS events can be controlled by properly sized automatic safety systems

Planning the research and development necessary for accelerator transmutation of waste, leading to integrated proof of performance testing

International Nuclear Information System (INIS)

Bennett, D.R.; Pasamehmetoglu, K.; Finck, P.; Pitcher, E.; Khalil, H.; Todosow, M.; Hill, R.; Van Tuyle, G.; Laidler, J.; Crawford, D.; Thomas, K.

2001-01-01

The Research and Development (R and D) Plan for the Accelerator Transmutation of Waste (ATW) Program has been developed for the Department of Energy, Office of Nuclear Energy (DOE/NE) to serve as a focus and progressional guide in developing critical transmutation technologies. It is intended that the Plan will serve as a logical reference considering all elements of an integrated accelerator-driven transmutation system, and will maximize the use of resources by identifying and prioritizing research, design, development and trade activities. The R and D Plan provides a structured framework for identifying and prioritizing activities leading to technically-justifiable integrated Proof of Performance testing within ten years and ultimate demonstration of Accelerator Transmutation of Waste (ATW). The Plan builds from the decision objectives specified for ATW, utilizes informational input from the ATW Roadmap and programmatic System Point Design efforts, and employs the knowledge and expertise provided by professionals familiar with ATW technologies. With the firm intent of understanding what, why and when information is needed, including critical interfaces, the Plan then develops a progressional strategy for developing ATW technologies with the use of a Technology Readiness Level (TRL) scale. The TRL approach is first used to develop a comprehensive, yet generic, listing of experimental, analytical and trade study activities critical to developing ATW technologies. Technology-specific and concept-specific aspects are then laid over the generic mapping to gage readiness levels. Prioritization criteria for reducing technical uncertainty, providing information to decision points, and levering off of international collaborations are then applied to focus analytical, experimental and trade activities. (author)

Comparison of energy expenditure between aquatic and overground treadmill walking in people post-stroke.

Science.gov (United States)

Jung, Taeyou; Ozaki, Yoshi; Lai, Byron; Vrongistinos, Konstantinos

2014-03-01

This study aimed to compare the cardiorespiratory responses between aquatic treadmill walking (ATW) and overground treadmill walking (OTW) in people with hemiparesis post-stroke. Eight participants post-stroke aged 58.5 ± 11.4 years and eight healthy adult controls aged 56.1 ± 8.6 years participated in a cross-sectional comparative study. Participants completed three 8-minute walking sessions separated by at least 72-hour rest. On the first visit, participants identified their comfortable walking speed on an aquatic and overground treadmill. The second and third visit consisted of either ATW or OTW at a matched speed. Oxygen consumption (VO2), carbon dioxide production (VCO2 ), minute ventilation (VE) and energy expenditure (EE) were measured at rest and during walking in both exercise modes. Mean steady-state cardiorespiratory responses during ATW showed a significant decrease compared with OTW at a matched speed. During ATW, mean VO2 values decreased by 39% in the stroke group and 21% in the control group, mean VCO2 values decreased by 42% in the stroke group and 30% in the control group, and mean EE decreased by 40% in the stroke group and 25% in the control group. Mean steady-state VE values and resting cardiorespiratory response values showed no significant change between the two conditions. This study demonstrated a decreased metabolic cost when ATW at matched speeds to that of OTW. Reduced metabolic cost during ATW may allow for longer durations of treadmill-induced gait training compared with OTW for improved outcomes. This knowledge may aid clinicians when prescribing aquatic treadmill exercise for people post-stroke with goals of improving gait and functional mobility. However, decreased metabolic cost during ATW suggests that to improve cardiovascular fitness, ATW may not be a time-efficient method of cardiovascular exercise for healthy adults and people post-stroke. Copyright © 2013 John Wiley & Sons, Ltd.

Accelerator transmutation of waste economics

International Nuclear Information System (INIS)

Krakowski, R.A.

1995-01-01

A parametric systems model of the accelerator transmutation of (nuclear) waste (ATW) is used to examine key system trade-offs and design drivers on the basis of unit costs. This model is applied primarily to a fluid-fuel blanket concept for an ATW that generates net electric power from the fissioning of spent commercial reactor fuel. An important goal of this study is the development of essential parametric trade-offs to aid in any future conceptual engineering design of an ATW that would burn spent commercial fuel and generate net electric power. As such, costing procedures and methodologies used to estimate and compare advanced nuclear power generation systems are applied. The cost of electricity required by an electrical power-generating ATW fueled with spent commercial fuels is generally found to be above that projected for other advanced fission power plants. The accelerator and the chemical plant equipment cost accounts are quantitatively identified as main cost drivers, with the capital cost of radio-frequency power dominating the former. Significant reductions of this cost differential are possible by increased blanket neutron multiplication, increased plant capacity, or increased thermal-to-electric conversion efficiency. The benefits of reduced long-lived fission products and spent commercial fuel actinides provided by the ATW approach translate into a less tangible source of revenue to be provided by a charge that must be levied on the client fission power plants being serviced. The main goal of this study, however, is not a direct cost comparison but is instead a quantitative determination of cost-based sensitivity of key cost drivers and operational modes for an ATW concept that would address the growing spent commercial fuel problem; parametric results presented focus on this goal, and a specific ATW ''straw man'' is given to achieve this main objective

Irradiated fuel behavior during reactivity initiated accidents in LWR's: Status of research and development studies in France

International Nuclear Information System (INIS)

Papin, J.; Merle, J.P.

1994-01-01

There is much interest in the nuclear industry concerning the ability of training simulators to adequately model severe accident conditions, specifically Anticipated Transient Without Scram (ATWS) events. The Pennsylvania Power and Light Co. has recently installed a new simulator which was provided by S3 Technologies. As part of the licensed operator training program, PP ampersand L provides training on Emergency Operating Procedures (EOPs). Since the ATWS event is challenging from both a computational and operational point of view, the Engineering Department was asked to benchmark the new simulator performance. The purpose of this benchmark was to ensure simulator fidelity with EOP basis calculations which are numerically more rigorous. Once acceptable simulator fidelity had been demonstrated, EOPs were evaluated to ensure they could be implemented by the operators. This paper examines the details of the new simulator response for ATWS events, and exposes the PP ampersand L ATWS procedures to further examination. The simulator benchmark was carried out using the PP ampersand L-developed SABRE code which has been benchmarked against plant data and industry accepted codes. For many ATWS scenarios, the new simulator, which is based upon first principles, provides preditions consistent with SABRE. Reactor power levels, consistent with SABRE results, are significantly higher than predicted by the old simulator, and containment pressurization occurs much more rapidly than previously simulated. Additionally, the new simulated reactor water level, pressure and power are far more responsive to perturbations than predicted by the old simulator. This responsiveness is consistent with SABRE predictions and has helped to define modifications to the ATWS emergency operating procedures. The modified procedures enhance the operators ability to respond to ATWS given the much more realistic reactor model

Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Mi; Kim, Ji Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of); Seok, Ho [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

2016-10-15

An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics.

Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram

International Nuclear Information System (INIS)

Choi, Sun Mi; Kim, Ji Hwan; Seok, Ho

2016-01-01

An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics

Considerations on reduction of risk from anticipated transients without scram in a regulatory perspective

International Nuclear Information System (INIS)

Ahn, S. H.; Oh, D. Y.; Kim, I. K.; Lee, S. H.

2002-01-01

ATWS (Anticipated Transients Without Scram) are anticipated operational occurrences accompanied by the failure of the reactor trip portion of the reactor trip system. ATWS accidents are an cause of concern because under certain postulated conditions they could lead to significant core damage including core melt and to the large release of radioactivity to the environment. In this study, considerations on reduction of risk from ATWS were discussed with examination of the technical background of 10CFR 50.62. Considering the recent trends of the extended core cycle and the power uprating, it is recognized that the moderator temperature coefficient can become less negative than to suppress the RCS overpressure followed by ATWS. Because the negative reactivity feedback is one inherent level of multiple defenses, the effect against the RCS overpressure needs to be assessed in detail

Accelerator-driven destruction of long-lived radioactive waste and energy production

International Nuclear Information System (INIS)

Schriber, S.O.

1997-01-01

Nuclear waste management involves many issues. ATW is an option that can assist a repository by enhancing its capability and thereby assist nuclear waste management. Technology advances and the recent release of liquid metal coolant information from Russia has had an enormous impact on the viability of an ATW system. It now appears economic with many repository enhancing attributes. In time, an ATW option added to present repository activities will provide the public with a nuclear fuel cycle that is acceptable from economic and environmental points of view

NCBI nr-aa BLAST: CBRC-AGAM-04-0047 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-AGAM-04-0047 ref|ZP_01612901.1| hypothetical protein ATW7_13618 [Alteromonadale...s bacterium TW-7] gb|EAW27830.1| hypothetical protein ATW7_13618 [Alteromonadales bacterium TW-7] ZP_01612901.1 4.0 33% ...

NCBI nr-aa BLAST: CBRC-CREM-01-1192 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-CREM-01-1192 ref|ZP_01614196.1| hypothetical protein ATW7_04619 [Alteromonadale...s bacterium TW-7] gb|EAW26551.1| hypothetical protein ATW7_04619 [Alteromonadales bacterium TW-7] ZP_01614196.1 0.046 22% ...

NCBI nr-aa BLAST: CBRC-CELE-04-0052 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-CELE-04-0052 ref|ZP_01614196.1| hypothetical protein ATW7_04619 [Alteromonadale...s bacterium TW-7] gb|EAW26551.1| hypothetical protein ATW7_04619 [Alteromonadales bacterium TW-7] ZP_01614196.1 0.64 21% ...

Anticipated transients without scram for light water reactors

International Nuclear Information System (INIS)

1978-12-01

In the first two volumes of this report, Anticipated Transients without Scram for Light Water Reactors NUREG-0460, dated April 1978, the NRC staff reviewed the information on this subject that had been developed in the past and evaluated the susceptibility of current nuclear plants to ATWS events using fault tree/event tree analysis techniques. Based on that evaluation, the staff concluded that some corrective measures were required to reduce the risk of severe consequences arising from possible ATWS events. Since the issuance of NUREG-0460, new safety and cost information has become available on ATWS. Also, new insights have been developed on the general subject of quantitative risk assessment. The purpose of this supplement to NUREG-0460 is to summarize the important additions to the information base and to propose a course of action from among a variety of alternatives for resolving the ATWS concern

Transient analysis on the SMART-P anticipated transients without scram

International Nuclear Information System (INIS)

Yang, S. H.; Bae, K. H.; Kim, H. C.; Zee, S. Q.

2005-01-01

Anticipated transients without scram (ATWS) are anticipated operational occurrences accompanied by a failure of an automatic reactor trip when required. Although the occurrence probability of the ATWS events is considerably low, these events can result in unacceptable consequences, i.e. the pressurization of the reactor coolant system (RCS) up to an unacceptable range and a core-melting situation. Therefore, the regulatory body requests the installation of a protection system against the ATWS events. According to the request, a diverse protection system (DPS) is installed in the SMART-P (System-integrated Modular Advanced ReacTor-Pilot). This paper presents the results of the transient analysis performed to identify the performance of the SMART-P against the ATWS. In the analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code is applied to identify the thermal hydraulic response of the RCS during the transients

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

International Nuclear Information System (INIS)

Dallman, R.J.; Gottula, R.C.; Holcomb, E.E.; Jouse, W.C.; Wagoner, S.R.; Wheatley, P.D.

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key Topic / Enhanced safety and operation excellence / Radiation protection

Energy Technology Data Exchange (ETDEWEB)

Baumann, Erik [AREVA GmbH, Erlangen (Germany). Radiation Protection; Bohnsted, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Safety Research for Nuclear Waste Management, Radiation Protection

2015-10-15

Summary report on the Focus Session 'Radiation Protection' of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 and 8 (2015) and will be covered in further issues of atw.

Nanoindentation and micro-mechanical fracture toughness of electrodeposited nanocrystalline Ni–W alloy films

International Nuclear Information System (INIS)

Armstrong, D.E.J.; Haseeb, A.S.M.A.; Roberts, S.G.; Wilkinson, A.J.; Bade, K.

2012-01-01

Nanocrystalline nickel–tungsten alloys have great potential in the fabrication of components for microelectromechanical systems. Here the fracture toughness of Ni–12.7 at.%W alloy micro-cantilever beams was investigated. Micro-cantilevers were fabricated by UV lithography and electrodeposition and notched by focused ion beam machining. Load was applied using a nanoindenter and fracture toughness was calculated from the fracture load. Fracture toughness of the Ni–12.7 at.%W was in the range of 1.49–5.14 MPa √m. This is higher than the fracture toughness of Si (another important microelectromechanical systems material), but considerably lower than that of electrodeposited nickel and other nickel based alloys. - Highlights: ► Micro-scale cantilevers manufactured by electro-deposition and focused ion beam machining. ► Nanoindenter used to perform micro-scale fracture test on Ni-13at%W micro-cantilevers. ► Calculation of fracture toughness of electrodeposited Ni-13at%W thin films. ► Fracture toughness values lower than that of nanocrystalline nickel.

Nanoindentation and micro-mechanical fracture toughness of electrodeposited nanocrystalline Ni-W alloy films

Energy Technology Data Exchange (ETDEWEB)

Armstrong, D.E.J., E-mail: david.armstrong@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Haseeb, A.S.M.A. [Department of Mechanical Engineering, University of Malaya, 50603 Kuala Lumpur (Malaysia); Roberts, S.G.; Wilkinson, A.J. [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Bade, K. [Institut fuer Mikrostrukturtechnik (IMT), Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2012-04-30

Nanocrystalline nickel-tungsten alloys have great potential in the fabrication of components for microelectromechanical systems. Here the fracture toughness of Ni-12.7 at.%W alloy micro-cantilever beams was investigated. Micro-cantilevers were fabricated by UV lithography and electrodeposition and notched by focused ion beam machining. Load was applied using a nanoindenter and fracture toughness was calculated from the fracture load. Fracture toughness of the Ni-12.7 at.%W was in the range of 1.49-5.14 MPa {radical}m. This is higher than the fracture toughness of Si (another important microelectromechanical systems material), but considerably lower than that of electrodeposited nickel and other nickel based alloys. - Highlights: Black-Right-Pointing-Pointer Micro-scale cantilevers manufactured by electro-deposition and focused ion beam machining. Black-Right-Pointing-Pointer Nanoindenter used to perform micro-scale fracture test on Ni-13at%W micro-cantilevers. Black-Right-Pointing-Pointer Calculation of fracture toughness of electrodeposited Ni-13at%W thin films. Black-Right-Pointing-Pointer Fracture toughness values lower than that of nanocrystalline nickel.

Study of anticipated transient without scram for PWR

International Nuclear Information System (INIS)

Pu Jilong.

1985-01-01

Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

Post-exercise hypotensive responses following an acute bout of aquatic and overground treadmill walking in people post-stroke: a pilot study.

Science.gov (United States)

Lai, Byron; Jeng, Brenda; Vrongistinos, Konstantinos; Jung, Taeyou

2015-06-01

The purpose of this study is to investigate the effects of a single-bout of aquatic treadmill walking (ATW) and overground treadmill walking (OTW) on the magnitude and duration of post-exercise ambulatory blood pressure (BP) in people post-stroke. Seven people post-stroke participated in a cross-sectional comparative study. BP was monitored for up to 9 hours after a 15-minute bout of ATW and OTW at approximately 70% of maximal oxygen consumption (VO2max), performed on separate days. Mean systolic and diastolic BP values were compared between both exercise conditions and a day without exercise (control). Three hours after OTW, mean SBP increased by 9% from pre-exercise baseline compared to a 3% decrease during the control day (P stroke can sustain sufficient walking intensities necessary to reduce BP following cardiovascular exercise. Also, these data suggest that ATW can elicit clinically meaningful reductions in DBP and night-time SBP. Thus, it is recommended for clinicians to consider ATW as a non-pharmaceutical means to regulate DBP and promote nighttime dipping of SBP in people post-stroke. However, caution is advised during the immediate hours after exercise, a period of possible BP inflation.

In silico Prediction and Wet Lab Validation of Arisaema tortuosum (Wall.) Schott Extracts as Antioxidant and Anti-breast Cancer Source: A Comparative Study.

Science.gov (United States)

Kant, Kamal; Lal, Uma Ranjan; Ghosh, Manik

2018-01-01

Globally, reactive oxygen species have served as an alarm predecessor toward pathogenesis of copious oxidative stress-related diseases. The researchers have turned their attention toward plant-derived herbal goods due to their promising therapeutic applications with minimal side effects. Arisaema tortuosum (Wall.) Schott (ATWS) is used in the traditional medicine since ancient years, but scientific assessments are relatively inadequate and need to be unlocked. Our aim was designed to validate the ATWS tuber and leaf extracts as an inhibitor of oxidative stress using computational approach. The reported chief chemical entities of ATWS were docked using Maestro 9.3 (Schrödinger, LLC, Cambridge, USA) tool and further ATWS extracts (tubers and leaves) were validated with 2,2'-diphenyl-1-picrylhydrazyl (DPPH), 2,2'-azino-bis (3-ethylbenzothiazoline-6-sulfonic acid) diammonium salt (ABTS), ferric-reducing ability of plasma (FRAP), and sulforhodamine B assays experimentally. In silico results showed notable binding affinity of ATWS phytoconstituents with the receptor (PDB: 3ERT). Experimentally, butanolic tuber fraction confirmed promising antioxidant potential (ABTS: IC 50 : 271.67 μg/ml; DPPH: IC 50 : 723.41 μg/ml) with a noteworthy amount of FRAP (195.96 μg/mg), total phenolic content (0.087 μg/mg), and total flavonoid content (7.5 μg/mg) while chloroform fraction (leaves) showed considerable reduction in the cell viability of MCF-7 cell line. The current findings may act as a precious tool to further unlock novel potential therapeutic agents against oxidative stress. Quercetin showed top.ranked glide score with notable binding toward 3ERT receptorAmong extracts, butanolic tubers confirmed as promising antioxidant with remarkable amount of TPC and TFCIn addition, chloroform fraction (leaves) revealed considerable decline in the cell viability of MCF-7 cell line. Abbreviations used: ATWS: Arisaema tortuosum (Wall.) Schott, DPPH: 2,2'-diphenyl-1-picrylhydrazyl

Separations technology development to support accelerator-driven transmutation concepts

International Nuclear Information System (INIS)

Venneri, F.; Arthur, E.; Bowman, C.

1996-01-01

This is the final report of a one-year Laboratory-Directed Research and Development (LDRD) Project at the Los Alamos National Laboratory (LANL). This project investigated separations technology development needed for accelerator-driven transmutation technology (ADTT) concepts, particularly those associated with plutonium disposition (accelerator-based conversion, ABC) and high-level radioactive waste transmutation (accelerator transmutation of waste, ATW). Specific focus areas included separations needed for preparation of feeds to ABC and ATW systems, for example from spent reactor fuel sources, those required within an ABC/ATW system for material recycle and recovery of key long-lived radionuclides for further transmutation, and those required for reuse and cleanup of molten fluoride salts. The project also featured beginning experimental development in areas associated with a small molten-salt test loop and exploratory centrifugal separations systems

Cladding failure probability modeling for risk evaluations of fast reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Kramer, J.M.

1987-01-01

This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

46{sup th} annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations enhanced safety and operation excellence

Energy Technology Data Exchange (ETDEWEB)

Lamm, Matthias [AREVA GmbH, Erlangen (Germany). R and D; Hollands, Thorsten [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Reactor Safety Research Div.

2016-01-15

Summary report on the Technical Sessions ''Know-how, New Build and Innovations'' and ''Operation and Safety of Nuclear Installations, Fuel SA: WASA-BOSS + CESAM'' of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 12 (2015) and will be covered in further issues of atw.

46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations / Enhanced safety and operation excellence

Energy Technology Data Exchange (ETDEWEB)

Schaffrath, Andreas; Scheuerer, Martina [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany); Oenneby Carina; Benjaminsson, Ulf [Westinghouse Electric Sweden AB, Vaesteraes (Sweden). Europe, Middle East and Africa (EMEA)

2015-11-15

Summary report on the Topical Sessions ''CFD Simulations for Reactor Safety Relevant Objectives '' and ''Fuel Management During the Last Cycles and Beyond'' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 10 (2015) and will be covered in further issues of atw.

Analysis of a main steam isolation valve closure anticipated transient without scram in a boiling water reactor

International Nuclear Information System (INIS)

Liaw, T.J.; Pan, C.; Chen, G.S.

1989-01-01

Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main stream isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4

Sensitivity of reactivity feedback due to core bowing in a metallic-fueled core

International Nuclear Information System (INIS)

Nakagawa, Masatoshi; Kawashima, Masatoshi; Endo, Hiroshi; Nishimura, Tomohiro

1991-01-01

A sensitivity study has been carried out on negative reactivity feedback caused by core bowing to assess the potential effectiveness of FBR passive safety features in regard to withstanding an anticipated transient without scram (ATWS). In the present study, an analysis has been carried to obtain the best material and geometrical conditions concerning the core restraint system out for several power to flow rates (P/F), up to 2.0 for a 300 MWe metallic-fueled core. From this study, it was clarified that the pad stiffness at an above core loading pads (ACLP) needs to be large enough to ensure negative reactivity feedback against ATWS. It was also clarified that there is an upper limit for the clearances between ducts at ACLP. A new concept, in regard to increasing the absolute value for negative reactivity feedback due to core bowing at ATWS, is proposed and discussed. (author)

Planning and reporting of Russian transmutation research projects within ISTC. Phase 1

Energy Technology Data Exchange (ETDEWEB)

Conde, H. [Uppsala Univ., (Sweden). Dept. of Neutron Research; Gudowski, W. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Reactor Technology; Liljenzin, J.O. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Mileikovsky, C. [Pully (Switzerland)

1997-02-01

The International Scientific and Technical Center (ISTC) in Moscow funds research of civil interest to counteract the risk of nuclear weapon proliferation. Recently, new technical concepts, Accelerator Transmutation of Nuclear Waste (ATW), have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The Russian experts are knowledgeable and well equipped for doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been proposed to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. The present report describes the back ground, the status and near term activities of a few ISTC projects of relevance for the ATW concept, which are planned with the participation of a Swedish reference group. 4 refs.

Cladding failure probability modeling for risk evaluations of fast reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Kramer, J.M.

1987-01-01

This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

Planning and reporting of Russian transmutation research projects within ISTC. Phase 1

International Nuclear Information System (INIS)

Conde, H.

1997-02-01

The International Scientific and Technical Center (ISTC) in Moscow funds research of civil interest to counteract the risk of nuclear weapon proliferation. Recently, new technical concepts, Accelerator Transmutation of Nuclear Waste (ATW), have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The Russian experts are knowledgeable and well equipped for doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been proposed to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. The present report describes the back ground, the status and near term activities of a few ISTC projects of relevance for the ATW concept, which are planned with the participation of a Swedish reference group. 4 refs

Criticality safety analysis of accelerator transmutation waste system

International Nuclear Information System (INIS)

Landeyro, P.A.; Cepraga, D.G.; Orazi, A.

1993-01-01

The Accelerator Transmutation Waste system (ATW) is under development at the Los Alamos National Laboratory. It consists of a particle accelerator producing a proton beam having an energy of 1.5 GeV. These particles are introduced into the upper part of a molten Pb-Bi column and they produce, by a spallation reaction, a high strength neutron flux, 1.0x10 16 n/(square centimeters sec). The neutrons enter a heavy water blanket where actinides and long-lived fission products circulate in vertical tubes. The goal of this research effort is to perform an independent verification of the feasibility of actinide burning in the ATW system. The work is divided into four tasks: a) production of an actinide and long-lived fission product cross section library from JEF 2.2; b) simulation, using MCNP and KENO IV Monte Carlo codes, of the ATW configurations existing in literature; c) validation of the cross sections by comparison of Keff and reaction rate results, calculated with MCNP and KENO IV, with experimental benchmarks and intercomparison between calculations of a PWR unit cell and the computations carried out with various codes and cross section libraries (NEACRF criticality working group data); d) simulation of the ATW configuration. The two first tasks are almost complete with excellent agreement between this study's results and those of Los Alamos

Best-estimate methodology for analysis of anticipated transients without scram in pressurized water reactors

International Nuclear Information System (INIS)

Rebollo, L.

1993-01-01

Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests

Anticipated transient without scram analysis of the simplified boiling water reactor following main steam isolation valve closure with boron injection

International Nuclear Information System (INIS)

Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

1996-01-01

The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system

AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 2

International Nuclear Information System (INIS)

Fischer, Klaus-Christian; Willschuetz, Hans-Georg; Wortmann, Birgit

2014-01-01

Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Thermo Dynamics and Fluid Dynamics: Experiments and Backfittings for the Improvement of Safety and Efficiency; - Safety of Nuclear Installations - Methods, Analyses, Results: In-Vessel Phenomena; Ex-Vessel Phenomena; - Standards and Regulations; Hazard and Safety Analysis; and Validation and Uncertainty Analysis. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 (2014) and will be covered in further issues of atw.

46th Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

International Nuclear Information System (INIS)

Fischer, Erwin

2015-01-01

Summary report on the following Topical Session of the 46 th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 3

Energy Technology Data Exchange (ETDEWEB)

Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Programm Nukleare Sicherheitsforschung (NUKLEAR); Mull, Thomas [AREVA GmbH, Erlangen (Germany). Nuclear Fusion, HTR and Transverse Issues (PTDH-G); Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

2015-01-15

Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Reactor Operation, Safety: Radiation Protection (Angelika Bohnstedt); - Competence, Innovation, Regulation: Fusion Technology - Optimisation Steps in the ITER Design (Thomas Mull); - Competence, Innovation, Regulation: Education, Expert Knowledge, Knowledge Transfer (Joerg Starflinger). The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 and 12 (2015) and will be covered in further issues of atw.

46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

Energy Technology Data Exchange (ETDEWEB)

Fischer, Erwin [E.ON Kernkraft GmbH, Global Unit Next Generation, Hannover (Germany)

2015-08-15

Summary report on the following Topical Session of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

2011 annual meeting on nuclear technology. Pt. 4. Topical sessions

International Nuclear Information System (INIS)

Schoenfelder, Christian; Dams, Wolfgang

2011-01-01

Summary report on the Topical Session of the Annual Conference on Nuclear Technology held in Berlin, 17 to 19 May 2011: - Nuclear Competence in Germany and Europe. The Topical Session: - Sodium Cooled Fast Reactors -- will be covered in a report in a further issue of atw. The reports on the Topical Sessions: - CFD-Simulations for Safety Relevant Tasks; and - Final Disposal: From Scientific Basis to Application; - Characteristics of a High Reliability Organization (HRO) Considering Experience Gained from Events at Nuclear Power Stations -- have been covered in atw 7, 8/9, and 10 (2011). (orig.)

ACCELERATOR TRANSMUTATION OF WASTE TECHNOLOGY AND IMPLEMENTATION SCENARIOS

International Nuclear Information System (INIS)

Beller, D.; Tuyle, G. van

2000-01-01

During 1999, the U.S. Department of Energy, in conjunction with its nuclear laboratories, a national steering committee, and a panel of world experts, developed a roadmap for research, development, demonstration, and deployment of Accelerator-driven Transmutation of Waste (ATW). The ATW concept that was examined in this roadmap study was based on that developed at the Los Alamos National Laboratory (LANL) during the 1990s. The reference deployment scenario in the Roadmap was developed to treat 86,300 tn (metric tonnes initial heavy metal) of spent nuclear fuel that will accumulate through 2035 from existing U.S. nuclear power plants (without license extensions). The disposition of this spent nuclear reactor fuel is an issue of national importance, as is disposition of spent fuel in other nations. The U.S. program for the disposition of this once-through fuel is focused to characterize a candidate site at Yucca Mountain, Nevada for a geological repository for spent fuel and high-level waste. The ATW concept is being examined in the U.S. because removal of plutonium minor actinides, and two very long-lived isotopes from the spent fuel can achieve some important objectives. These objectives include near-elimination of plutonium, reduction of the inventory and mobility of long-lived radionuclides in the repository, and use of the remaining energy content of the spent fuel to produce power. The long-lived radionuclides iodine and technetium have roughly one million year half-lives, and they are candidates for transport into the environment via movement of ground water. The scientists and engineers who contributed to the Roadmap Study determined that the ATW is affordable, doable, and its deployment would support all the objectives. We report the status of the U.S. ATW program describe baseline and alternate technologies, and discuss deployment scenarios to support the existing U.S. nuclear capability and/or future growth with a variety of new fuel cycles

Materials compatibility and corrosion issues for accelerator transmutation of waste

International Nuclear Information System (INIS)

Staudhammer, K.

1992-08-01

The need to understand the materials issues in an accelerator transmutation of waste (ATW) system is essential. This report focuses on the spallation container material, as this material is exposed to some of the most crucial environmental conditions of simultaneous radiation and corrosion in the system. The most severe design being considered is that of liquid lead. In previous investigations of lead compatibility with materials, the chemistry of the system was derived solely from the corrosion products; however, in an ATW system, the chemistry of the lead changes not only with the derived corrosion products of the material being tested but also with the buildup of the daughter production with time. Daughter production builds up and introduces elements that may have a great effect on the corrosion activity of the liquid lead. Consequently, data on liquid lead compatibility can be regarded only as a guide and must be reevaluated when particular daughter products are added. This report is intended to be a response to specific materials issues and concerns expressed by the ATW design working group and addresses the compatibility/corrosion concerns

Disposition of nuclear waste using subcritical accelerator-driven systems

International Nuclear Information System (INIS)

Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

1998-01-01

Studies have shown that the repository long-term radiological risk is from the long-lived transuranics and the fission products Tc-99 and I-129, thermal loading concerns arise mainly form the short-lived fission products Sr-90 and Cs-137. In relation to the disposition of nuclear waste, ATW is expected to accomplish the following: (1) destroy over 99.9% of the actinides; (2) destroy over 99.9% of the Tc and I; (3) separate Sr and Cs (short half-life isotopes); (4) separate uranium; (5) produce electricity. In the ATW concept, spent fuel would be shipped to a ATW site where the plutonium, other transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their only pass through the facility. This approach contrasts with the present-day reprocessing practices in Europe and Japan, during which high purity plutonium is produced and used in the fabrication of fresh mixed-oxide fuel (MOX) that is shipped off-site for use in light water reactors

Decommissioning and dismantling of the Rossendorf Isotope Production

International Nuclear Information System (INIS)

Grahnert, Thomas

2016-01-01

After just over 40 years of production operation 2000, the operation of the Rossendorf Isotope Production was finally stopped. In the last few years of production already sections of the Rossendorf Isotope Production have been decommissioned. With the end of the isotope production the decommissioning of the entire complex started. In the two-part report, the decommissioning and dismantling of the Rossendorf Isotope production is presented. In part 1 (atw 5/2016) mainly the authorisation procedures and the realised decommissioning concept are presented. Part 2 (atw 6/2016) deals with special selected aspects of the implementation of the decommissioning programme.

Analyses of anticipated transient without scram events in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Chun, Ji Han; Kim, Soo Hyoung; Yang, Soo Hyung; Bae, Kyoo Hwan

2012-01-01

SMART is a small integral reactor, which was developed at KAERI and acquired standard design approval in 2012. SMART works like a pressurized light water reactor in principle though it is more compact than loop type large commercial reactors. ATWS(Anticipated Transient Without Scram) event is an AOO(Anticipated Operational Occurrence) where RPS fails to trip the reactor when requested. SMART incorporated a DPS(diverse protection system) to protect the reactor system when RPS(reactor protection system) fails to trip the reactor. The results of transient analyses show that DPS in SMART effectively mitigates the consequence of ATWS

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.

1987-01-01

The uncertainty in predicting the effectiveness of inherent shutdown (ISD) in innovative designs results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require ISD; (2) the approximation of representing all such ATWS events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. In this summary, the methodology and associated results of work used to establish probabilities of failure of inherent shutdown of innovative LMRs to the unprotected loss-of-flow (LOF) accident are discussed

RPV housed ATWS poison tank

International Nuclear Information System (INIS)

Oosterkamp, W.J.

1992-01-01

This patent describes a boiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core and an upper steam dome connected to a steam outlet in the RPV. The improvement comprises: a pressurized vessel disposed in the steam dome containing a neutron poison effective for inactivating the core and a first line for assaying the poison which first line runs to the outside of the RPV, the vessel being vented to the steam dome to pressurize the poison contained therein, the vessel being connected by a second line terminating beneath the core, the second line containing a valve which is actuable to release the poison through the line upon its actuation

Transmutation of radioactive waste: Effect on the nuclear fuel cycle

International Nuclear Information System (INIS)

Rasmussen, N.C.; Pigford, T.H.

1997-01-01

A committee of the National Research Council reviewed three concepts for transmuting radionuclides recovered from the chemical reprocessing of commercial light-water-reactor (LWR) fuel: LWR transmutation reactors fueled with recycled actinides, advanced liquid-metal reactors (ALMRs), and accelerator-driven subcritical reactors for transmutation of waste (ATW). The concepts were evaluated in terms of: (1) the extent to which waste disposal would benefit from transmutation, (2) time required to reduce the total inventory of radionuclides in the waste and fuel cycle, (3) the complexity of the overall transmutation system, (4) the extent of new development required, and (5) institutional and economic problems of operating such systems. Transmutation could affect geologic disposal of waste by reducing the inventory of transuranics (TRUs), fission products, and other radionuclides in the waste. Reducing the inventory of transuranics does not necessarily affect radiation doses to people who use contaminated ground water if the dissolution rate of transuranics in waste is controlled by elemental solubilities. However, reducing inventories of Am and Pu would decrease potential hazards from human intrusion. The likelihood for underground nuclear criticality would also be reduced. The long-lived fission products Tc-99, I-129, Cs-135 and others typically contribute most to the long-term radiation doses to future populations who use contaminated water from the repository. Their transmutation requires thermal or epithermal neutrons, readily available in LWR and ATW transmutors. ALMR and LWR transmutors would require several hundred years to reduce the total transuranic inventory by even a factor of 10 at constant electric power, and thousands of years for a hundred-fold reduction. For the same electrical power, the ATW could reduce total transuranic inventory about tenfold more rapidly, because of its very high thermal-neutron flux. However, extremely low process losses would be

Fuel rod behaviour during transients

International Nuclear Information System (INIS)

Bilsby, C.F.; Haste, T.J.; Garlick, A.; Cameron, R.F.

1982-04-01

The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

Pyrochemical separations technologies envisioned for the U.S. accelerator transmutation of waste system

International Nuclear Information System (INIS)

Laidler, J. J.

2000-01-01

A program has been initiated for the purpose of developing the chemical separations technologies necessary to support a large Accelerator Transmutation of Waste (ATW) system capable of dealing with the projected inventory of spent fuel from the commercial nuclear power stations in the United States. The baseline process selected combines aqueous and pyrochemical processes to enable the efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel. The diversity of processing methods was chosen for both technical and economic factors. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system

Boiling water reactor containment modeling and analysis at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Holcomb, E.E. III; Wilson, G.E.

1984-01-01

Under the auspices of the United States Nuclear Regulatory Commission, severe accidents are being studied at the Idaho National Engineering Laboratory. The boiling water reactor (BWR) studies have focused on postulated anticipated transients without scram (ATWS) accidents which might contribute to severe core damage or containment failure. A summary of the containment studies is presented in the context of the analytical tools (codes) used, typical transient simulation results and the need for prototypical containment data. All of these are related to current and future analytical capabilities. It is shown that torus temperatures during the ATWS depart from limiting conditions for BWR T-quencher operation, outside of which stable steam condensation has not been proven

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

Dynamic simulation of a sodium-cooled fast reactor power plant

Energy Technology Data Exchange (ETDEWEB)

Shinaishin, M.A.M.

1976-08-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

Safety Analysis for Key Design Features of KALIMER-600 Design Concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

2007-02-15

This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

Status report for anticipated transients without scram for Combustion Engineering reactors

International Nuclear Information System (INIS)

1975-01-01

The NRC staff review of Combustion ATWS analyses included the anticipated transients expected to occur, the initial conditions and system parameters assumed in the analyses, the reliability of systems, the analytical techniques, the results of transient analysis of ATWS events and the design of the Reactor Protection System. Using the requirements of WASH-1270 as a guideline, the staff reviewed each relevant aspect of the Combustion model and analysis. The discussion of anticipated transients is presented, and the initial conditions, system parameters, and operating systems assumed in the analyses of these transients are discussed. The analytical techniques and computer programs are reviewed. An independent calculation conducted by the staff using the RELAP-3B code to determine the pressure within the reactor coolant pressure boundary during a complete loss of main feedwater ATWS event is described. A set of standard problems is defined for all pressurized water reactor vendors and the Regulatory staff to insure acceptability of computer codes used in all systems transient analyses. The model for calculating water discharge through primary valves is described. The comparison of the Combustion analyses to the requirements of WASH-1270 is presented. Certain outstanding issues are identified which require that Combustion or the applicant provide additional information or modify existing designs

Dynamic simulation of a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Shinaishin, M.A.M.

1976-08-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant

Digital implementation of AMSACs at Harris and Robinson plants

International Nuclear Information System (INIS)

Burjorjee, D.; Stepps, D.

1988-01-01

The Code of Federal Regulations was altered in July 1984 to include a section on Requirements for Reduction of Risk from Anticipated Transients Without Scram Events for Light Water Cooled Nuclear Power Plants. For pressurized water reactors the code required equipment diverse from the reactor trip system to automatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine trip under conditions indicative of an anticipated transient without scram (ATWS). The equipment in question is called ATWS mitigation system actuation circuitry (AMSAC). The AMSACs for Carolina Power and Light Company's Shearon Harris and Robinson power plants have been designed and built by Atomic Energy of Canada Limited (AECL) from commercially available components to meet stringent reliability requirements and minimize operational burdens

BWR stability analysis

International Nuclear Information System (INIS)

Valtonen, K.

1990-01-01

The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

Evaluation of BWROG EPG level/power control strategy for Vermont Yankee

International Nuclear Information System (INIS)

Chandola, V.; Robichaud, J.D.

1987-01-01

The current Boiling Water Reactor Owner's Group (BWROG) emergency procedure guidelines (EPGs) direct reactor operators to manually lower the reactor pressure vessel (RPV) water level to the top of active fuel (TAF) during an anticipated transient without scram (ATWS) event. Lowering the water level reduces the core inlet flow, thereby reducing core power. However, reducing water level is contrary to current operator training, which requires that normal RPV water level be maintained to assure core cooling. In addition, the indicated water level near TAF using cold calibrated level instrumentation may not be reliable, which could potentially result in uncovering the core. This paper evaluates the EPGs' level/power control strategy for the Vermont Yankee plant and proposes alternative to the BWROG guidelines as applied to ATWS response

An investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a boiling water reactor anticipated transient without scram

International Nuclear Information System (INIS)

Peterson, C.E.; Chexal, V.K.; Gose, G.C.; Hentzen, R.D.; Layman, W.H.

1985-01-01

Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel

Anticipated transients without scram

International Nuclear Information System (INIS)

Lellouche, G.S.

1980-01-01

This article discusses in various degrees of depth the publications WASH-1270, WASH-1400, and NUREG-0460, and has as its purpose a description of the technical work done by Electric Power Research Institute (EPRI) personnel and its contractors on the subject of anticipated transients without scram (ATWS). It demonstrates the close relation between the probability of scram failure derived from historical scram data and that derived from the use of component data in a model of a system (the so-called synthesis method), such as was done in WASH-1400. The inherent conservatism of these models is demonstrated by showing that they predict significantly more events than have in fact occurred and that such models still predict scram failure probabilities low enough to make ATWS an insignificant contributor to accident risk

Screening and analysis of beyond design basis accident of 49-2 SPR

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

2015-01-01

The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Wade, D.C.

1988-01-01

The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. This paper discusses the work being done to address each of these contributing areas, identifies the design and research approaches being used at Argonne National Laboratory to reducing the key contributions to uncertainties in inherent shutdown, and presents results. The conditional probabilities (given ATWS initiation) of achieving temperatures capable of defeating inherent shutdown are shown to range from /approximately/0.1% to negligible for current designs

A study on the boron injection initiation temperature curve of BWR

International Nuclear Information System (INIS)

Wang, S.-J.; Chien, C.-S.; Fann, S.-Y.; Chiang, S.-C.

2007-01-01

Boron injection initiation temperature (BIIT) provides important information for the safe shutdown of the reactor using boron injection system during anticipated transient without scram (ATWS). The purpose of this paper is to study BIIT curve of boiling water reactor owners' group (BWROG). The unreasonable and non-conservative parts of BIIT are pointed out and suggested modifications are made. The starting reactor power of BIIT is increased in order to meet the actual application. The lower limit of suppression pool temperature of BIIT is revised for conservative operation during ATWS conditions. Analysis of the effects of maximum temperature capacity of the suppression chamber and concentration of boron in standby liquid control tank shows that BIIT is decreased by adopting a more conservative value of maximum temperature capacity of the suppression chamber. Consequently, early boron injection is anticipated. For system with automatic boron injection system, BIIT is not required

The Los Alamos accelerator driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

International Nuclear Information System (INIS)

Venneri, F.; Williamson, M.A.; Ning, L.

1997-01-01

The studies carried out in the frame of the Accelerator Driven Transmutation Technology (ADTT) program developed at Los Alamos in order to solve the nuclear waste problem and to build a new generation of safer and non-proliferant nuclear power plants, are presented

Dynamic simulation of a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Shinaishin, M.A.M.

1976-01-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is dealt with. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. In addition to the usual assumption of lumped parameters, uniform heat transfer and point kinetics (prompt jump) have been the main approximations in this and other simulators (see below). Two different transport-delay models have also been installed in all simulators. The simulators were constructed using the DARE-P System, developed by the Electrical Engineering Department at the University of Arizona

Disposition of nuclear waste using subcritical accelerator-driven systems

International Nuclear Information System (INIS)

Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

1998-01-01

Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power

Nuclear power plants: 2013 atw compact statistics

Energy Technology Data Exchange (ETDEWEB)

Anon.

2014-03-15

At the end of 2013, nuclear power plants were available for energy supply in 31 countries of the world. A total of 437 nuclear power plants were in operation with an aggregate gross power of approx. 393 GWe and an aggregate net power, respectively, of 372 GWe. This means that the number was unchanged compared to the previous year's number on 31 December 2012. The available gross power of nuclear power plants increased by approx. 2 GWe from 2012 to the end of 2013. In total 4 nuclear generating units were commissioned in 2013 in China (+2) and in the Republic Korea (+1). 6 nuclear generating units were decommissioned in 2013. Four units in the U.S.A. (-4) were shut down due to economical reasons. In Canada (-2) the operation status of 2 units was changed from long-term shutdown to permanently shutdown. 70 nuclear generating units with an aggregate gross power of approx. 73 GWe, were under construction in 15 countries end of 2013. New or continued projects are notified from (in brackets: number of new projects) China (+3), Belarus (+1), Rep. of Korea (+1) and the United Arab Emirates (+1). Some 115 new nuclear power plants are in the concrete project design, planning and licensing phases worldwide; on some of them, contracts have already been awarded. Another units are in their preliminary project phases. (orig.)

Nuclear power plants: 2009 atw compact statistics

International Nuclear Information System (INIS)

Anon.

2010-01-01

At the turn of 2009/2010, nuclear power plants were available for energy supply in 30 countries of the world. A total of 437 nuclear power plants, which is one plant less than at the 2008/2009 turn, were in operation with an aggregate gross power of approx. 391 GWe and an aggregate net power, respectively, of 371 GWe. The available gross power of nuclear power plants did not changed noticeably from 2008 to the end of 2009. In total 2 nuclear generating units were commissioned in 2009. One NPP started operation in India and one in Japan. Three nuclear generating units in Japan (2) und Lithuania (1) were decomissioned in 2009. 52 nuclear generating units, i.e. 10 plants more than at the end of 2008, with an aggregate gross power of approx. 51 GWe, were under construction in 14 countries end of 2009. New or continued projects are notified from (number of new projects): China (+9), Russia (1), and South Korea (1). Some 84 new nuclear power plants are in the concrete project design, planning and licensing phases worldwide; on some of them, contracts have already been awarded. Another units are in their preliminary project phases. (orig.)

Nuclear power plants: 2013 atw compact statistics

International Nuclear Information System (INIS)

Anon.

2014-01-01

At the end of 2013, nuclear power plants were available for energy supply in 31 countries of the world. A total of 437 nuclear power plants were in operation with an aggregate gross power of approx. 393 GWe and an aggregate net power, respectively, of 372 GWe. This means that the number was unchanged compared to the previous year's number on 31 December 2012. The available gross power of nuclear power plants increased by approx. 2 GWe from 2012 to the end of 2013. In total 4 nuclear generating units were commissioned in 2013 in China (+2) and in the Republic Korea (+1). 6 nuclear generating units were decommissioned in 2013. Four units in the U.S.A. (-4) were shut down due to economical reasons. In Canada (-2) the operation status of 2 units was changed from long-term shutdown to permanently shutdown. 70 nuclear generating units with an aggregate gross power of approx. 73 GWe, were under construction in 15 countries end of 2013. New or continued projects are notified from (in brackets: number of new projects) China (+3), Belarus (+1), Rep. of Korea (+1) and the United Arab Emirates (+1). Some 115 new nuclear power plants are in the concrete project design, planning and licensing phases worldwide; on some of them, contracts have already been awarded. Another units are in their preliminary project phases. (orig.)

Nuclear power plants: 2004 atw compact statistics

International Nuclear Information System (INIS)

Anon.

2005-01-01

In late 2004, nuclear power plants were available for power supply or were under construction in 32 countries worldwide. A total of 441 nuclear power plants, i.e. two plants more than in late 2003, were in operation with an aggregate gross power of approx. 386 GWe and an aggregate net power, respectively, of 362 GWe, in 31 countries. The available capacity of nuclear power plants increased by approx. 5 GWe as a result of the additions by the six units newly commissioned: Hamaoka 5 (Japan), Ulchin 6 (Korea), Kalinin 3 (Russia), Khmelnitski 2 (Ukraine), Qinshan II-2 (People's Republic of China), and Rowno 4 (Ukraine). In addition, unit 3 of the Bruce A nuclear power plant in Canada with a power of 825 MWe was restarted after an outage of many years. Contrary to earlier plans, a recommissioning program was initiated for the Bruce A-1 and A-2 units, which are also down at present. Five plants were decommissioned for good in 2004; Chapelcross 1 to 4 with 50 MWe each in the United Kingdom, and Ignalina 1 with 1 300 MWe in Lithuania. 22 nuclear generating units with an aggregate gross power of 19 GWe in nine countries were under construction in late 2004. In India, construction work was started on a new project, the 500 MWe PFBR prototype fast breeder reactor. In France, the EDF utility announced its intention to build an EPR on the Flamanville site beginning in 2007. (orig.)

Nuclear power plants: 2005 atw compact statistics

International Nuclear Information System (INIS)

Anon.

2006-01-01

Nuclear power plants were available for power supply and under construction, respectively, in 32 countries of the world as per end of 2005. A total of 444 nuclear power plants, i.e. three plants more than at the end of 2004, with an aggregate gross power of approx. 389 GWe and an aggregate net power of 370 GWe, respectively, were in operation in 31 countries. The available capacity of nuclear power plants increased by some 4,5 GWe as a result of the capacities added by the four newly commissioned units of Higashidori 1 (Japan), Shika 2 (Japan), Tarapur 4 (India), and Tianwan 1 (China). In addition, unit A-1 of the Pickering nuclear power station in Canada, with 825 MWe, was restarted after a downtime of several years. Two plants were decommissioned for good in 2005: Obrigheim in Germany, and Barsebaeck 2 in Sweden. 23 nuclear generating units, i.e. one unit more than in late 2004, with an aggregate gross power of approx. 19 GWe were still under construction in nine countries by late 2005. In Pakistan, construction of a new project, Chasnupp 2, was started; in China, construction was begun of two units, Lingao Phase 2, units 3 and 4, and in Japan, the Shimane 3 generating unit is being built. (orig.)

Nuclear power plants: 2008 atw compact statistics

International Nuclear Information System (INIS)

Anon.

2009-01-01

At the turn of 2008/2009, nuclear power plants were available for energy supply in 31 countries of the world. A total of 438 nuclear power plants, which is one plant less than at the 2007/2008 turn, were in operation with an aggregate gross power of approx. 393 GWe and an aggregate net power, respectively, of 372 GWe. The available gross power of nuclear power plants didn't changed noticeabely from 2007 to the end of 2008. No nuclear generating unit was commissioned in 2008. One nuclear generating unit in the Slovak Republic was decomissioned in 2008. 42 nuclear generating units, i.e. 10 plants more than at the end of 2007, with an aggregate gross power of approx. 38 GWe, were under construction in 14 countries end of 2008. New or continued projects are notified from (in brackets: number of new projects): Bulgaria (2), China (5), South Korea (2), Russia (1), and the Slovak Republic (2). Some 80 new nuclear power plants are in the concrete project design, planning and licensing phases worldwide; on some of them, contracts have already been awarded. Another approximately 120 units are in their preliminary project phases. (orig.)

47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key topic / Outstanding know-how and sustainable innovations

Energy Technology Data Exchange (ETDEWEB)

Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Bereich Reaktorsicherheitsforschung

2016-07-15

Summary report on the Technical Session: ''Reactor Physics, Thermo- and Fluid-Dynamics'' of the 47th Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 will be covered in further issues of atw.

Electronic properties of interfaces produced by silicon wafer hydrophilic bonding

Energy Technology Data Exchange (ETDEWEB)

Trushin, Maxim

2011-07-15

The thesis presents the results of the investigations of electronic properties and defect states of dislocation networks (DNs) in silicon produced by wafers direct bonding technique. A new insight into the understanding of their very attractive properties was succeeded due to the usage of a new, recently developed silicon wafer direct bonding technique, allowing to create regular dislocation networks with predefined dislocation types and densities. Samples for the investigations were prepared by hydrophilic bonding of p-type Si (100) wafers with same small misorientation tilt angle ({proportional_to}0.5 ), but with four different twist misorientation angles Atw (being of < , 3 , 6 and 30 , respectively), thus giving rise to the different DN microstructure on every particular sample. The main experimental approach of this work was the measurements of current and capacitance of Schottky diodes prepared on the samples which contained the dislocation network at a depth that allowed one to realize all capabilities of different methods of space charge region spectroscopy (such as CV/IV, DLTS, ITS, etc.). The key tasks for the investigations were specified as the exploration of the DN-related gap states, their variations with gradually increasing twist angle Atw, investigation of the electrical field impact on the carrier emission from the dislocation-related states, as well as the establishing of the correlation between the electrical (DLTS), optical (photoluminescence PL) and structural (TEM) properties of DNs. The most important conclusions drawn from the experimental investigations and theoretical calculations can be formulated as follows: - DLTS measurements have revealed a great difference in the electronic structure of small-angle (SA) and large-angle (LA) bonded interfaces: dominating shallow level and a set of 6-7 deep levels were found in SA-samples with Atw of 1 and 3 , whereas the prevalent deep levels - in LA-samples with Atw of 6 and 30 . The critical twist

Preliminary analysis of the induced structural radioactivity inventory of the base-case aqueous accelerator transmutation of waste reactor concept

International Nuclear Information System (INIS)

Bezdecny, J.A.; Vance, K.M.; Henderson, D.L.

1995-01-01

The purpose of the Los Alamos National Laboratory Accelerator Transmutation of (Nuclear) Waste (ATW) project is the substantial reduction in volume of long-lived high-level radioactive waste of the US in a safe and energy-efficient manner. An evaluation of the ATW concept has four aspects: material balance, energy balance, performance, and cost. An evaluation of the material balance compares the amount of long-lived high-level waste transmuted with the amount and type, of waste created in the process. One component of the material balance is the activation of structural materials over the lifetime of the transmutation reactor. A preliminary radioactivity and radioactive mass balance analysis has been performed on four structure regions of the reaction chamber: the tungsten target, the lead annulus, six tubing materials carrying the actinide slurry, and five reaction vessel structural materials. The amount of radioactive material remaining after a 100-yr cooling period for the base-case ATW was found to be 338 kg of radionuclides. The bulk of this material (313 kg) was generated in the zirconium-niobium (Zr-Nb) actinide tubing material. Replacement of the Zr-Nb tubing material with one of the alternative tubing materials analyzed would significantly reduce the short- and long-term radioactive mass produced. The alternative vessel material Al-6061 alloys, Tenelon, HT-9, and 2 1/4 Cr-1 Mo and the alternative actinide tubing materials Al-6061 alloy, carbon-carbon matrix, silicon carbide, and Ti-6 Al-4 V qualify for shallow land burial. Alternative disposal options for the base-case structural material Type 304L stainless steel and the actinide tubing material Zr-Nb will need to be considered as neither qualifies for shallow land burial

47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key Topics / Enhanced safety and operation excellence

Energy Technology Data Exchange (ETDEWEB)

Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Programm Nukleare Entsorgung, Sicherheit und Strahlenforschung (NUSAFE); Baumann, Erik [AREVA GmbH, Erlangen (Germany). Radiation Protection

2016-12-15

Summary report on the Key Topic 'Enhanced Safety and Operation Excellence' Focus Session 'Radiation Protection' of the 47{sup th} Annual Meeting on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 have been and will be covered in further issues of atw.

Accelerator-driven transmutation of high-level waste from the defense and commercial sectors

International Nuclear Information System (INIS)

Bowman, C.; Arthur, E.; Beard, C.

1996-01-01

This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The major goal has been to develop accelerator transmutation of waste (ATW) system designs that will thoroughly and rapidly transmute nuclear waste, including plutonium from dismantled weapons and spent reactor fuel, while generating useful electrical power and without producing a long-lived radioactive waste stream. We have identified and quantified the unique qualities of subcritical nuclear systems and their capabilities in bringing about the complete destruction of plutonium. Although the 1191 subcritical systems involved in our most effective designs radically depart from traditional nuclear reactor concepts, they are based on extrapolations of existing technologies. Overall, care was taken to retain the highly desired features that nuclear technology has developed over the years within a conservative design envelope. We believe that the ATW systems designed in this project will enable almost complete destruction of nuclear waste (conversion to stable species) at a faster rate and without many of the safety concerns associated with the possible reactor approaches

47{sup th} Annual conference on nuclear technology (AMNT 2016). Key topics / Outstanding know-how and sustainable innovations - enhanced safety and operation excellence

Energy Technology Data Exchange (ETDEWEB)

Raetzke, Christian [CONLAR - Consulting on Nuclear Law, Licensing and Regulation, Leipzig (Germany); Fischer, Erwin [PreussenElektra GmbH, Hannover (Germany). Management Board; Mohrbach, Ludger [VGB PowerTech e.V., Essen (Germany). Competence Center ' ' Nuclear Power Plants' '

2016-08-15

Summary report on the Key Topics ''Outstanding Know-How and Sustainable Innovations'' and ''Enhanced Safety and Operation Excellence'' of the 47{sup th} Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 will be covered in further issues of atw.

On the ATW-concepts: ITP approach and opportunities

Science.gov (United States)

Simonenko, V. A.; Grebyonkin, K. F.

1995-09-01

It is discussed the interest of Russian Federal Nuclear Center-Institute of Technical Physics at Chelyabinsk-70 in the research of Accelerator Driven Technologies applications for radioactive waste transmutation, cumulated actinides burning, energy production. The ITP background and opportunities for this research are presented. It is shown the ITP possibilities for testing and experimental development of Accelerator Driven Technologies.

On the ATW-concepts: ITP approach and opportunities

Energy Technology Data Exchange (ETDEWEB)

Simonenko, V.A.; Gregyonkin, K.F. [Institute of Technical Physics, Chelyabinsk (Russian Federation)

1995-10-01

It is discussed the interest of Russian Federal Nuclear Center - Institute of Technical Physics at Chelyabinsk-70 in the research of Accelerator Driven Technologies applications for radioactive waste transmutation, cumulated actinides burning, energy production. The ITP background and opportunities for this research are presented. It is shown the ITP possibilities for testing and experimental development of Accelerator Driven Technologies.

76 FR 78872 - National Emission Standards for Hazardous Air Pollutants for Wool Fiberglass Manufacturing

Science.gov (United States)

2011-12-20

... effort to accommodate all speakers that arrive and register before 8 p.m. The EPA's Web site for the... preferences on speaking times may not be able to be fulfilled. If you will require the service of a translator..., including lists of speakers, will be posted on the EPA's Web site at http://www.epa.gov/ttn/atw/rrisk/rtrpg...

AMNT 2014. Key Topic: Reactor operation, safety - report. Pt. 1

International Nuclear Information System (INIS)

Schaffrath, Andreas

2014-01-01

Summary report on one session of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Safety of Nuclear Installations - Methods, Analysis, Results: Backfittings for the Improvement of Safety and Efficiency. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' will be covered in further issues of atw.

Kinematics of two-phase mixture level motion in BWR pressure vessels

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

1985-01-01

A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key topic / Outstanding know-how and sustainable innovations

Energy Technology Data Exchange (ETDEWEB)

Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

2016-11-15

Summary report on the Key Topic ''Outstanding Know-How and Sustainable Innovations'' Technical Session ''Reactor Physics, Thermo, and Fluid Dynamics'' of the 47th Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 have been and will be covered in further issues of atw.

Fabrication of Ni-5 at. %W Long Tapes with CeO₂ Buffer Layer by Reel-to-Reel Method

DEFF Research Database (Denmark)

Ma, Lin; Tian, Hui; Yue, Zhao

2015-01-01

A 10-m-long homemade textured Ni-5at.%W (Ni5W) long tape with a CeO2 buffer layer has been prepared successfully by means of rolling-assisted biaxially textured substrate (RABiTS) route followed by a chemical solution deposition method in a reel-to-reel manner. Globally, the Ni5W substrate and CeO2...

47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key Topics / Enhanced safety and operation excellence and decommissioning experience and Waste management solutions

Energy Technology Data Exchange (ETDEWEB)

Salnikova, Tatiana [AREVA GmbH, Erlangen (Germany); Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-10-15

Summary report on the Key Topics ''Enhanced Safety and Operation Excellence'' and ''Decommissioning Experience and Waste Management Solutions'' of the 47{sup th} Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 have been and will be covered in further issues of atw.

Development of 4S and related technologies. (3) Statistical evaluation of safety performance of 4S on ULOF event

International Nuclear Information System (INIS)

Ishii, Kyoko; Matsumiya, Hisato; Horie, Hideki; Miyagi, Kazumi

2009-01-01

The purpose of this work is to evaluate quantitatively and statistically the safety performance of Super-Safe, Small, and Simple reactor (4S) by analyzing with ARGO code, a plant dynamics code for a sodium-cooled fast reactor. In this evaluation, an Anticipated Transient Without Scram (ATWS) is assumed, and an Unprotected Loss of Flow (ULOF) event is selected as a typical ATWS case. After a metric concerned with safety design is defined as performance factor a Phenomena Identification Ranking Table (PIRT) is produced in order to select the plausible phenomena that affect the metric. Then a sensitivity analysis is performed for the parameters related to the selected plausible phenomena. Finally the metric is evaluated with statistical methods whether it satisfies the given safety acceptance criteria. The result is as follows: The Cumulative Damage Fraction (CDF) for the cladding is defined as a metric, and the statistical estimation of the one-sided upper tolerance limit of 95 percent probability at a 95 percent confidence level in CDF is within the safety acceptance criterion; CDF < 0.1. The result shows that the 4S safety performance is acceptable in the ULOF event. (author)

Nuclear power plants 1995 - a world survey

International Nuclear Information System (INIS)

Anon.

1996-01-01

The atw Statistics Report compiled by atw lists 428 nuclear power plants with 363 397 gross MWe in operation in 30 countries in late 1995. Another 62 units with 55 180 gross MWe were under construction in 18 countries. This adds up to a total of 490 units with an aggregate 418 577 MWe. In the course of 1995 four units in four countries started commercial operation. In the survey of electricity generation in 1995 for which no information was made available from China and Kasachstan, a total of 417 nuclear power plants were covered. In the year under review they generated an aggregate 2 282 614 GWH, which is 3.4% more than in the previous year. The highest nuclear generation again was recorded in the USA with 705 771 GWh, followed by France with 377 021 GWh. The Grohnde power station in Germany attained the maximum annual production figure of 11 359 GWh. The survey includes nine tables indicating the generating performance of each nuclear power plant, the development of electricity generation in nuclear plants, and status of nuclear power plants at the end of 1995 arranged by countries, types of reactors, and reactor manufacturers. (orig.) [de

Preliminary investigation of actinide and xenon reactivity effects in accelerator transmutation of waste high-flux systems

International Nuclear Information System (INIS)

Olson, K.R.; Henderson, D.L.

1995-01-01

The possibility of an unstable positive reactivity growth in an accelerator transmutation of waste (ATW)-type high-flux system is investigated. While it has always been clear that xenon is an important actor in the reactivity response of a system to flux changes, it has been suggested that in very high thermal flux transuranic burning systems, a positive, unstable reactivity growth could be caused by the actinides alone. Initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately. The maximum change in reactivity after a flux change caused by the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or startup. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response caused by the actinides. The capabilities and applications of both the current actinide model and the xenon model are discussed. Finally, the need for a complete dynamic model for the high-flux fluid-fueled ATW system is addressed

AMNT 2014. Key Topic: Reactor operation, safety - report. Pt. 1

Energy Technology Data Exchange (ETDEWEB)

Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany). Forschungszentrum

2014-10-15

Summary report on one session of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Safety of Nuclear Installations - Methods, Analysis, Results: Backfittings for the Improvement of Safety and Efficiency. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' will be covered in further issues of atw.

Fuel rod response to BWR power oscillations during anticipated transient without scram

International Nuclear Information System (INIS)

Cunningham, M.; Scott, H.

1998-01-01

The US NRC is examining fuel behaviour during a postulated BWR anticipated transient without scram (ATWS) with power oscillations to determine if current regulatory criteria are adequate. Currently, the 280 cal/g limit for RIAs is used to show that coolable geometry is maintained and pressure pulses are avoided during ATWSs. Two specific questions have now been raised about the continued use of the 280 cal/g value. First, this value was derived from energy deposition values whereas the regulatory requirements are written in terms of fuel enthalpy. The second is that fuel rod rupture with fuel dispersal has been observed in RIA tests with high bum-up fuel rods having energy deposition values well below the current limit. However, the BWR ATWS power oscillation transient is slower than a RIA power pulse, thus reducing the likelihood of failure. Therefore questions about the adequacy of the 280 cal/g limit do not necessarily imply unacceptable fuel damage occurring during such power oscillations and there is no immediate safety concern. The reported analysis, using the FRAPTRAN transient fuel rod analysis code, was thus undertaken to determine if further investigation might be appropriate and with the intention of starting some discussions about the issue. There was a comment that a limit of 100 cal/g fuel enthalpy had been mentioned following the scoping calculations but that perhaps enthalpy was not the main concern in an ATWS. It was also observed that cladding stresses are lower than in all RIA. The question was what really is the main concern. It was replied that the main concern was a question of maintaining a coolable geometry i.e. not loosing fuel particles out of the rod. And it was agreed that enthalpy may not be the important issue, rather that it previously had been used as the parameter and so had been considered. Confirmation of this presently being an evaluation and not a regulatory concern was sought and provided, it being pointed out that the NRC

Transmutation of long-lived nuclides

International Nuclear Information System (INIS)

Liang Tongxiang; Tang Chunhe

2003-01-01

Partitioning and transmutation of long-lived nuclides have profound benefits for economic development, global political stability and the environment. This technology would reduce nuclear waste disposal requirements, prevent proliferation and eliminate a major hurdle to the development of nuclear power. This paper reviews the advanced fuel cycle process and development of ATW in the world, and some suggestions about the R and D of nuclear power in China are proposed

2008 annual meeting on nuclear technology. Pt. 1. Section reports; JAHRESTAGUNG KERNTECHNIK 2008. T. 1. Sektionsberichte

Energy Technology Data Exchange (ETDEWEB)

Dagan, Ron; Sanchez Espinoza, Victor Hugo [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit; Rohde, U.; Kliem, Soeren [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany); Faber, Wolfgang; Berlepsch, Thilo v.; Spann, Holger [E.ON Kernkraft GmbH, Hannover (Germany); Schaffrath, Andreas [TUEV Nord SysTec GmbH und Co. KG, Hamburg (Germany); Schubert, Bernd [Vattenfall Europe Nuclear Energy GmbH, Hamburg (Germany); Rieger, Udo [Vattenfall Nuclear Energy GmbH, Hamburg (Germany); Christ,, Bernhard G. [NUKEM Technologies GmbH, Alzenau (Germany); Gulden, Werner [Fusion for Energy, Barcelona (Spain); Bogusch, Edgar [AREVA NP GmbH, Erlangen (Germany)

2008-08-15

Summary report on these 5 - out of 11 - Sections of the Annual Conference on Nuclear Technology held in Hamburg on May 27-29, 2008: - Reactor Physics and Methods of Calculation - Thermodynamics and Fluid Dynamics - Safety of Nuclear Installations - Methods, Analysis, Results - Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage - Fusion Technology. Other Sections will be covered in reports in further issues of atw. (orig.)

2008 annual meeting on nuclear technology. Pt. 1. Section reports

International Nuclear Information System (INIS)

Dagan, Ron; Sanchez Espinoza, Victor Hugo; Faber, Wolfgang; Berlepsch, Thilo v.; Spann, Holger; Schaffrath, Andreas; Schubert, Bernd; Rieger, Udo; Christ, Bernhard G.; Gulden, Werner; Bogusch, Edgar

2008-01-01

Summary report on these 5 - out of 11 - Sections of the Annual Conference on Nuclear Technology held in Hamburg on May 27-29, 2008: - Reactor Physics and Methods of Calculation - Thermodynamics and Fluid Dynamics - Safety of Nuclear Installations - Methods, Analysis, Results - Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage - Fusion Technology. Other Sections will be covered in reports in further issues of atw. (orig.)

Transient analysis for resolving safety issues

International Nuclear Information System (INIS)

Chao, J.; Layman, W.

1987-01-01

The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

Theoretical investigations of fuel behavior during LOCA and ATWS

International Nuclear Information System (INIS)

Meyder, R.; Unger, H.

1976-01-01

The program system SSYST has been improved. The results of the SSYST-FRAP comparison calculations showed good agreement. In both programs, for instance, ballooning at the hottest spot occurs almost at the same time (appr 8 s). The calculation of the experiments of IRB on ballooning led also to a qualitative good agreement of experimental and theoretical results. The parameters in Nortons creep law are quantitatively not yet satisfactory. Gas gap flow equalizes axial pressure difference already at small gaps. The method of 'Moment Matching' for the statistical analysis needs considerably less computer time than 'Monte Carlo' method, and differs only slightly in expected values and variances. (orig./RW) [de

ATWS: a reappraisal. Part 3. Frequency of anticipated transients

International Nuclear Information System (INIS)

McClymont, A.S.; Poehlman, B.W.

1982-01-01

This document is the first revision of Part 3 of the EPRI study of the anticipated transients without scram question. This revision includes an update of events at nuclear power plants which had led to fast reactor shutdowns (scrams). The purpose of this document is to present the nuclear power plant operating experience, reflecting the frequency of these events identified by their principal characteristics

Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

International Nuclear Information System (INIS)

Liaw, T.J.; Pan, C.; Chen, G.S.

1989-01-01

Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

Planning and reporting of Russian transmutation research projects within ISTC. Phase 2

Energy Technology Data Exchange (ETDEWEB)

Conde, H. [Uppsala Univ. (Sweden). Dept. of Neutron Research; Gudowski, W. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Reactor and Neutron Physics; Liljenzin, J.O. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Mileikovsky, C. [Pully (Switzerland)

1998-11-01

The present report about phase 2 of the SKI project on Planning and Reporting of Russian Transmutation Research Projects within ISTC is an update of the information given in the SKI report no 97:15 (Feb 1997) about phase 1 of the same project. The background information is partly repeated in the present report to avoid that the reader has to go back to the report of Phase 1 for information about the basis for the project. USA, EU, Japan, Republic of Korea and Norway are at present supporting the International Scientific and Technical Center (ISTC) in Moscow. The Centre gives funds to research projects of civilian interest to former nuclear weapon laboratories to counteract the risk of nuclear weapon proliferation by the emigration of former USSR technical and scientific experts to `border countries` which are aiming towards the development of nuclear weapons. Before Sweden and Finland entered the EU, both countries gave national support to ISTC, in the case of Sweden 4 MUSD. Some of the projects which were funded by the Swedish national support to ISTC are still in progress. Nuclear technical concepts (i.e. Accelerator Transmutation of Nuclear Waste, ATW) have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The named Russian experts are knowledgeable and well equipped of doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been initiated, and further ones have been proposed, to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. A similar centre STCU (The Scientific and Technical Centre of the Ukraine) has been set up in Kiev. Sweden has been active in promoting this Centre, which is supported by USA, Japan, Canada and recently also by EU. The present report describes the

Planning and reporting of Russian transmutation research projects within ISTC. Phase 2

International Nuclear Information System (INIS)

Conde, H.

1998-11-01

The present report about phase 2 of the SKI project on Planning and Reporting of Russian Transmutation Research Projects within ISTC is an update of the information given in the SKI report no 97:15 (Feb 1997) about phase 1 of the same project. The background information is partly repeated in the present report to avoid that the reader has to go back to the report of Phase 1 for information about the basis for the project. USA, EU, Japan, Republic of Korea and Norway are at present supporting the International Scientific and Technical Center (ISTC) in Moscow. The Centre gives funds to research projects of civilian interest to former nuclear weapon laboratories to counteract the risk of nuclear weapon proliferation by the emigration of former USSR technical and scientific experts to 'border countries' which are aiming towards the development of nuclear weapons. Before Sweden and Finland entered the EU, both countries gave national support to ISTC, in the case of Sweden 4 MUSD. Some of the projects which were funded by the Swedish national support to ISTC are still in progress. Nuclear technical concepts (i.e. Accelerator Transmutation of Nuclear Waste, ATW) have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The named Russian experts are knowledgeable and well equipped of doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been initiated, and further ones have been proposed, to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. A similar centre STCU (The Scientific and Technical Centre of the Ukraine) has been set up in Kiev. Sweden has been active in promoting this Centre, which is supported by USA, Japan, Canada and recently also by EU. The present report describes the

European nuclear features. Interview: Commissioner Piebalgs (DG TREN)

International Nuclear Information System (INIS)

2005-01-01

Fifth issue of the European Nuclear Features. A joint publication of atw, Nuclear Espana, and Revue Generale Nucleaire. Contents: Frontier research in the EU Scientific Council of the European Research Council announced proactive safety management strategies; Pebble Bed Modular Reactor; New experiences applying methodologies for control room I and C modernization; Experts from Taiwan fulfil their training in Madrid; Ensa's activities in the Asian Commercial Nuclear Power Market; Clearance survey approach for scrap metals from NPP. (orig.)

Material for 258 atom bombs disappeared?

International Nuclear Information System (INIS)

Gruemm, H.

1988-01-01

In a report published in the news magazine, 'Der Spiegel', it was said that IAEA safeguards obviously had failed, for large amounts of fissile material had disappeared, which could be turned into 258 atomic bombs. The article in this issue of atw by the former Deputy Director General with the IAEA Safeguards Division sketches the background to the assertions made by 'Der Spiegel' and presents an overview of the inspection and verification methods employed by IAEA. (orig./HP) [de

Human reliability analysis for venting a BWR Mark I during a severe accident

International Nuclear Information System (INIS)

Nelson, W.R.; Blackman, H.S.

1986-01-01

A Human Reliability Analysis (HRA) was performed for the operator actions necessary to achieve containment venting for the Peach Bottom Atomic Power Station. This study was funded by the United States Nuclear Regulatory Commission (USNRC) and performed by the Idaho National Engineering Laboratory (INEL). The goal of the analysis was to estimate Human Error Probabilities (HEPs) to determine the likelihood that operators would fail to complete the venting process. The analysis was performed for two generic accident sequences: anticipated transient without scram (ATWS) and station blackout. Two major methods were used to estimate the HEPs: Technique for Human Error rate Prediction (THERP) and Success Likelihood Index Methodology (SLIM). For the ATWS scenarios analyzed, the calculated HEPs ranged from 0.23 to 0.35, depending on the number of vent paths that are required to reduce the containment pressure. It should be noted that the confidence bounds around these HEPs are large, However, even when considering the large confidence range, the failure probabilities are larger than what is typical for normal operator actions. For station blackout, the HEP is 1.0, resulting from the dangerous environmental conditions that are present, assuming that plant management would not deliberately expose personnel to a potentially fatal environment. These results are based on the analysis of draft procedures for containment venting. It is probable that careful revision of the procedures could reduce the human error probabilities

Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

International Nuclear Information System (INIS)

Ball, S.J.; Nypaver, D.J.

1994-01-01

An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ''operator'' involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor's fuel or metallic components

Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

Energy Technology Data Exchange (ETDEWEB)

Ball, S.J.; Nypaver, D.J.

1994-06-01

An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.

Why have we stopped research on liquid centrifugal separation

International Nuclear Information System (INIS)

Li, N.

1996-01-01

Using high-temperature high-speed liquid centrifuges for lanthanides and actinides separation was originally proposed as a physical separation method in the Los Alamos ADTT/ATW concept [C. Bowman, LA-UR-92-1065 (1992)]. The authors investigated centrifugal separation in a concerted effort of experiments, theoretical analysis and numerical simulations. They discovered that owing to the ionic-composition-dependence of the sedimentation coefficients for the fission products and actinides, separation by grouping of molecular densities would not work in general in the molten salt environment. Alternatively the lanthanides and actinides could be transferred to a liquid metal carrier (e.g. bismuth) via reductive extraction and then separated by liquid centrifuges, but the material and technical challenges are severe. Meanwhile the authors have established that the reductive extraction procedure itself can be used for desired separations. Unlike conventional aqueous-based reprocessing technologies, reductive extraction separation uses only reagent (Li) that reconstitutes carrier salts (LiF-BeF 2 ) and a processing medium (Bi) that can be continuously recycled and reused, with a nearly-pure fission products waste stream. The processing units are compact and reliable, and can be built at relatively low cost while maintaining high throughput. Therefore the research effort on developing liquid centrifuges for separations in ADTT/ATW was terminated in late 1995. This paper will discuss the various aspects involved in reaching this decision

Safety analysis and neutronics of accelerator-driven transmutation of wastes with concurrent energy production. Annual report for the year 1996

International Nuclear Information System (INIS)

Gudowski, W.; Thedeen, T.; Moeller, E.

1997-02-01

The research activities have significantly expanded compared to the earlier period and were during 1996 concentrated on the following major objectives: ATW system studies, simulations, optimization and design of spallation targets, benchmarking of the calculational tools in the frame of the IAEA coordinated research project, development of the computer codes for ADS and some experimental activities on the subcritical reactor MASURCA. Under 1996 a very extensive international collaboration network has been further developed and many collaborative research projects have been launched. 31 refs

Accelerator-driven thermal fission systems may provide energy supply advantages

International Nuclear Information System (INIS)

Linford, R.K.

1992-01-01

This presentation discusses the energy supply advantages of using accelerator-driven thermal fission systems. Energy supply issues as related to cost, fuel supply stability, environmental impact, and safety are reviewed. It is concluded that the Los Alamos Accelerator Transmutation of Waste (ATW) concept, discussed here, has the following advantages: improved safety in the form of low inventory and subcriticality; reduced high-level radioactive waste management timescales for both fission products and actinides; and a very long-term fuel supply requiring no enrichment

Results of recent LOFT experiments

International Nuclear Information System (INIS)

Leach, L.P.; Hanson, D.J.; Batt, D.L.

1982-01-01

Five experiments were performed in the Loss-of-Fluid Test (LOFT) facility during the past year. The experiments conducted spanned a wide range of potential accident scenarios, including large and small break loss-of-coolant accidents (LOCAs), control rod withdrawal accidents, uncontrolled boron dilution, and anticipated transients without scram (ATWS). This summary describes these experiments and presents results available from the experiments and experiment prediction calculations. A brief overview is given for the remaining experiment planned in the LOFT Program

Common cause failure analysis of the rodded scram system of the Arkansas Nuclear One-Unit 1 Plant

International Nuclear Information System (INIS)

Montague, D.F.; Campbell, D.J.; Flanagan, G.F.

1986-10-01

This study demonstrates the use of a formal method for common cause failure analysis in a reliability analysis of the Arkansas Nuclear One - Unit 1 rodded scram system. The scram system failure of interest is loss of capability of the system to shut the reactor down when required. The results of this analysis support the ATWS program sponsored by the US Nuclear Regulatory Commission. The methods used in this analysis support the NRC's Risk Methods Integration and Evaluation Program (RMIEP)

Emittance growth and halo formation in charge-dominated beams

Energy Technology Data Exchange (ETDEWEB)

Bondarev, B.I.; Durkin, A.P.; Murin, B.P. [Moscow Radiotechnical Institute (Russian Federation)] [and others

1995-10-01

The optimization of high-current high-energy linacs against the low beam loss requirement is not straightforward or well-codified. Outlying particle losses at the 10{sup {minus}5} up to 10{sup {minus}8} level might have only a small effect on the rms properties of the beam, and thus the total beam size must be constantly kept under observation. RMS-physics has gained wide-spread acceptance as a necessary design tool, but its sufficiency is an issue for ATW/ABC accelerators.

Controlled energy generation from nuclear fusion. 60th year atw

Energy Technology Data Exchange (ETDEWEB)

Weiss, Georg [Pintsch Bamag AG, Frankfurt am Main (Germany)

2015-02-15

Prospects increase, that with a controlled process of nuclear fusion one day an additional nuclear energy source will be commercially exploitable. In what follows, scientific principles according to the most recent research will be presented. Since approximately 30 years we are aware of the fact, that energy in form of light and heat provided by the sun and other fixed stars since over four billions years resulted from reactions of atomic nuclei. A series of such reactions became known which are considered for 'thermonuclear' processes, for example the carbon cycle by Bethe, where hydrogen is converted into helium. Most of the reflections and experiments dealt until 1938 with the reaction between nuclei of light elements. The possibility of splitting heavy nuclei was not anticipated. Its discovery by Hahn and Strassmann was a complete surprise - so to speak a rash reaction to release energy at the end of the element row. This 'way out' captured the interest of nuclear physicist for more than a decade. Only today, by starting to construct big nuclear power plants - only today, being able to assess the possibilities and limitations of this technology, the idea of energy generation through nuclear fusion steps into the foreground of nuclear research.

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of ''inherent shutdown'' is emphasized in the approach to the design of innovative, small pool-type liquid metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram (ATWS) for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower (TOP) events in evolving metal and oxide innovative designs

Transmutation of Isotopes --- Ecological and Energy Production Aspects

Science.gov (United States)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional

German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

International Nuclear Information System (INIS)

Velkov, K.

2002-01-01

Appendix 1: The Safety Analysis Report (S.A.R.) is presented from 3 Handbooks - ECC Handbook (LOCA), Plant Dynamics Handbook (Transients incl. ATWS), and Core Design Handbook. The first one Conceived as Living handbook, Basis for design, catalogue of transients, specifications and licensing. Handbook contains LOCA in primary system, it contains also core damage analysis, and description of codes, description of essential plant data and code input data. The second one consists of Basis for design, commissioning, operation, and catalogue of transients, specifications and licensing, as well as specified operation, disturbed operation, incidents, non-LOCA, SS-procedures and Code description. The third book consists of Reactivity balance and reactivity coefficients, efficiency of shutdown systems. Calculation of burn up cycle, power density distribution, and critical boron concentration. Also Codes used, as SAV79A standard analysis methodology including FASER for nuclear data generation, MEDIUM and PANBOX for static and transient core calculations. Appendix 2: The three TUEV (Technical Inspection Agencies) responsible for the three individual plants of type KONVOI: TUEV Bayern for ISAR-2, TUV-Hanover for KKE, TUEV-Stuttgart for GKN-2 and GRS performed the safety assessment. TUV-Bayern for disturbance and failure of secondary heat sink without loss of coolant (failure of main heat sink, erroneous operation of valves in MS and in FW system, failure of MFW supply), long term LONOP, performance of selected SBLOCA analyses. TUV Hanover for disturbances due to failure of MCPs, short term LONOP, damages of SG tubes incl. SGTR, performance of selected LOCA analyses (blowdown phase of LBLOCA). TUV-Stuttgart for breaks and leaks in MS and FW system with and without leaks in SG tubes. GRS for ATWS, sub-cooling transients due to disturbances on secondary side, initial and boundary conditions for transients with opening of pressurizer valves with and without stuck-open, most of the

Water level measurement uncertainty during BWR instability

International Nuclear Information System (INIS)

Torok, R.C.; Derbidge, T.C.; Healzer, J.M.

1994-01-01

This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by ppercase RETRAN for an ATWS/instability event.The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument. ((orig.))

Upgrade of Control and Protection System of the Ignalina Nuclear Power Plant Units 1 and 2

International Nuclear Information System (INIS)

Wright, Ronald E.; Fletcher, Norman; Sidnev, Victor E.; Bickel, John H.; Vianello, Aldo; Pearsall, Raymond D.

2003-01-01

The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy

Five MW Nuclear Heating Reactor

International Nuclear Information System (INIS)

Zhang Dafang; Dong Duo; Su Qingshan

1997-01-01

The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy Technology (INET) has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation, and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection and environmental impacts and so on were also obtained at the same time. All of these results demonstrate the design of the NHR-5 is successful. (author). 9 refs, 11 figs, 5 tabs

A five MW nuclear heating reactor

International Nuclear Information System (INIS)

Zhang Dafang; Don Duo; Su Quingshan

1997-01-01

The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy and Technology (INET) and has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection, and environmental impacts and so on, were also obtained at the same time. All of these demonstrate that the design of NHR-5 is successful. (author)

Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

International Nuclear Information System (INIS)

Varde, P. V.; Lee, D. Y.; Han, J. B.

2003-03-01

A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

Reactor safety and radiation protection. Draft of the BMU deparmental budget 16 of the 1998 German federal budget

International Nuclear Information System (INIS)

Anon.

1997-01-01

The expenditures earmarked for reactor safety and radiation protection in the 1998 budget of the German Federal Ministry for the Environment, Nature Conservation, and Reactor Safety (BMU) total DM 101 million. The expenditures of the German Federal Office for Radiation Protection (BfS) are to amount to a total of DM 579 million. These are the figures included in departmental budget 16 of the 1998 federal budget, which was discussed by the Federal Parliament in September 1997. The atw compilation singles out a number of significant items of the departmental budget. (orig.) [de

Calculations of radiation damage in target, container and window materials for spallation neutron sources

International Nuclear Information System (INIS)

Wechsler, M.S.; Mansur, L.K.

1996-01-01

Radiation damage in target, container, and window materials for spallation neutron sources is am important factor in the design of target stations for accelerator-driver transmutation technologies. Calculations are described that use the LAHET and SPECTER codes to obtain displacement and helium production rates in tungsten, 316 stainless steel, and Inconel 718, which are major target, container, and window materials, respectively. Results are compared for the three materials, based on neutron spectra for NSNS and ATW spallation neutron sources, where the neutron fluxes are normalized to give the same flux of neutrons of all energies

Protozoa and their bacterial prey colonize sterile soil fast

DEFF Research Database (Denmark)

Altenburger, Andreas; Ekelund, Flemming; Jacobsen, Carsten Suhr

2010-01-01

We know little about the ability of protozoa to colonize soils, including their successional patterns. To elucidate this issue, we investigated in which order different protozoan morpho-types colonize sterile soil. We used sterilized soils with different carbon content, and exposed them to the at......We know little about the ability of protozoa to colonize soils, including their successional patterns. To elucidate this issue, we investigated in which order different protozoan morpho-types colonize sterile soil. We used sterilized soils with different carbon content, and exposed them...

Five MW Nuclear Heating Reactor

Energy Technology Data Exchange (ETDEWEB)

Dafang, Zhang; Duo, Dong; Qingshan, Su [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy Technology (INET) has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation, and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection and environmental impacts and so on were also obtained at the same time. All of these results demonstrate the design of the NHR-5 is successful. (author). 9 refs, 11 figs, 5 tabs.

Probability and consequences of severe reactor accidents. 60th year atw

Energy Technology Data Exchange (ETDEWEB)

Rasmussen, Norman Carl [Massachusetts Institute of Technology (MIT), Cambridge, MA (United States). Dept. of Nuclear Engineering

2015-06-15

The study carried out on behalf of former USAEC (United States Atomic Energy Commission) led by Prof. Rasmussen and published in reworked form as WASH 1400 by the USNRC (United States Nuclear Regulatory Commission) in 1975, assessed in 3,300 pages the risks that can be deducted from severe accidents in nuclear power plants. The results, often quoted and criticised, were so far the most conclusive statements to this question. In his lecture at the reactor meeting in 1976, Prof. Rasmussen tried to trace back the conclusion of the results to the question: Is the use of larger nuclear power plants, in accordance to experiences and calculations so far, acceptable? His risk assessment, related to American power plants and cites, on behalf of the BMI is currently evaluated by the IRS together with the LRA on specific occurrences within the Federal Republic of Germany.

Study of the burning capability of the Los Alamos ATW system

Energy Technology Data Exchange (ETDEWEB)

Landeyro, P.A. [ENEA, Roma (Italy); Buccafurni, A.; Orazi, A. [ANPA, Roma (Italy)

1995-10-01

The aim of calculations is to evaluate the evolution of the infinate multiplication factor (k{sub inf}) during the irradiation of minor actinides, High Level Waste (HLWL) and Plutonium. The most important results are independently verified with Monte Carlo calculations. The relative importance of the main parameters affecting the k{sub inf} was investigated by performing calculations with several minor actinide and plutonium concentrations as well as different {sup 238}U decontamination factors for HLW. The merit figure value for minor actinide alone, considering a constant neutron flux indicates that the best results are reached for minor actinide concentration equal to PWR spent fuel. The best plutonium burning results are obtained for a concentration (50.23 g/l) equal to the half of PWR spent fuel one. The simulations lead to two different reactor concepts: one for HLW burning and the other for plutonium burning purposes. To burn the HLW the most suitable reactor is an homogeneous one. This kind of reactor can effectively be utilised to burn minor actinide in low concentration (namely the PWR spent fuel). On the other hand an heterogeneous reactor with channels filled by all actinides present in PWR spent fuel with the exclusion of U isotopes with a concentration of 50 g/l can be studied.

New nuclear power plants in Europe 1984. Pt 2

International Nuclear Information System (INIS)

Anon.

1984-01-01

The atw report on new nuclear power plants in Europe contains both a survey of the Federal Republic of Germany, which was pubslished in the April 1984 issue, and an overview of the nuclear power plant situation in 26 European countries including the Soviet Union and six other CMEA countries. Also this year's review includes specific status reports, complete with technical information, about all nuclear generating units under construction, in the project and concrete planning phases. The fifteen nuclear power plants newly commissioned in Europe since last year's atw report was published are covered in a similar way. Moreover, introductory summaries describe the plants in operation in each country and their 1983 electricity generation. A general introduction provides an outlook on developments in Western and Eastern Europe. The total number of nuclear power plants now in operation and under construction in Europe is 300 units with an aggregate gross 219, 320 MWe. Of these, 185 units are located in Western Europe, 115 in Eastern Europe. The units currently in operation of this total are 180 with 103,978 MWe in sixteen European countries; of these, 126 units with 74,869 MWe are run in eleven West European countries, 54 units with an aggregate 29,109 MWe in five East European countries. Of the 120 nuclear generating units at present under construction with an aggregate 115,342 MWe in fifteen European countries, 59 units with 63,442 MWe are located in eight West European, 61 units with 51,900 MWe in seven East European countries. (orig./UA) [de

Nuclear questions

International Nuclear Information System (INIS)

Hohlfeld, W.

1977-01-01

This brochure 'nuclear problems' deals with the attitude of the protestant church in the region around the northern Elbe towards further quantitative economic growth, esp. nuclear energy, with the following essays: preaching the Gospel in an environment in danger: the Christian occident and the problems of the third world, facing the problems of exhausted supplies, the role of the prophet, problem of environment - a problem of theology, the political dimension, against ATW, signal Brokdorf, strange effects (defense of the church from unqualified teachings by non-professionals), Christian liberty, church and nuclear energy, violence and robes. (HP) [de

Study on depressurization measurements and effect in PWR

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

Implementation of new regulations on nuclear powered plant design and operation raise new design and management requirement for plants, and the operational plants also need accident management to enhance the reactor operation safety. Thus, for sake of reducing risk of high-pressure and mitigating the consequence, depressurization is a measure carried out to reduce primary pressure. With SCDAP/RELAP5 this paper studies the depressurization measurements and effect factors in pressurized water reactor under the important severe accident sequences induced by very small break lost of coolant accident (VSBLOCA), anticipated transient without scram (ATWS) and station blackout (SBO) plus auxiliary feedwater failure. (author)

RF system considerations for accelerator production of tritium and the transmutation of nuclear waste

International Nuclear Information System (INIS)

Tallerico, P.J.; Lynch, M.T.

1993-01-01

RF driven proton accelerators for the transmutation of nuclear waste (ATW) or for the production of tritium (APT) require unprecedented amounts of CW RF power at UHF frequencies. For both systems, the baseline design is for 246 MW at 700 MHz and 8,5 MW at 350 MHz. The main technical challenges are how to design and build such a large system so that it has excellent reliability, high efficiency, and reasonable capital cost. The issues associated with the selection of the RF amplifier and the sizes of the power supplies are emphasized in this paper

Special MAFIA postprocessors for the analysis of RF structures

International Nuclear Information System (INIS)

Browman, M.J.

1992-01-01

This paper describes three stand-alone programs that use the electromagnetic fields generated by the MAFIA 2.04 codes to analyze radio-frequency (RF) cavities. Illustrations are provided that show how these codes are used to analyze the effect of the coupling slots on the electric and magnetic fields of the linacs for the APLE Prototype Experiment (APEX) and the Advanced Free-Electron Laser (AFEL); verify the Panofsky-Wenzel theorem for a high-energy deflecting cavity proposed for the Accelerator Transmutation of Waste (ATW) project; and study the effectiveness of that deflecting cavity. (Author) 10 refs., 4 figs

Special MAFIA postprocessors for the analysis of rf structures

International Nuclear Information System (INIS)

Browman, M.J.

1992-01-01

This paper describes three stand-alone programs that use the electromagnetic fields generated by the MAFIA 2.04 codes to analyze radio-frequency (RF) cavities. Illustrations are provided that show how these codes are used to (1) analyze the effect of the coupling slots on the electric and magnetic fields of the linacs for the APLE Prototype Experiment (APEX) and the Advanced Free-Electron Laser (AFEL), (2) verify the Panofsky-Wenzel theorem for a high-energy deflecting cavity proposed for the Accelerator Transmutation of Waste (ATW) project, and (3) study the effectiveness of that deflecting cavity

On the safety of nuclear installations in the Soviet Union

International Nuclear Information System (INIS)

Anon.

1989-01-01

The cooperation agreements between authorities and industries of the Soviet Union and West Germany now are gaining shape in practice. In this context, the framework conditions are of great interest that govern the realisation of the extensive nuclear energy programme of the Soviet Union. The chairman of the State Commission established in 1984 for supervision of nuclear installations and guidance on safety-engineering enhancement of nuclear power plant in the USSR has been interviewed by atw on topics of organisations, measures and regulatory activities in the field of reactor safety and radiation protection. The interview is given in full. (orig.) [de

Response surface methodology approach for structural reliability analysis: An outline of typical applications performed at CEC-JRC, Ispra

International Nuclear Information System (INIS)

Lucia, A.C.

1982-01-01

The paper presents the main results of the work carried out at JRC-Ispra for the study of specific problems posed by the application of the response surface methodology to the exploration of structural and nuclear reactor safety codes. Some relevant studies have been achieved: assessment of structure behaviours in the case of seismic occurrences; determination of the probability of coherent blockage in LWR fuel elements due to LOCA occurrence; analysis of ATWS consequences in PWR reactors by means of an ALMOD code; analysis of the first wall for an experimental fusion reactor by means of the Bersafe code. (orig.)

www.kernenergie.de - nuclear power has a German Internet address

International Nuclear Information System (INIS)

Anon.

2000-01-01

www.kernenergie.de is the address on the worldwide web under which the German nuclear organizations, Deutsches Atomforum (DAtF), Informationskreis Kernenergie (IK), and Kerntechnische Gesellschaft (KTG) as well as atw - internationale Zeitschrift fuer Kernenergie, and INFORUM can be reached. Extensive sources of information, discussions, on-line dictionaries, computer codes, dynamic web pages, digital documents and multimedia offerings can be called up via the portal under the individual web sites. In this way, www.kernenergie.de provides a comprehensive and up-to-date background of information about nuclear power and adjacent topics in the digital worldwide web. (orig.) [de

Mitigation of Severe Accident Consequences Using Inherent Safety Principles

International Nuclear Information System (INIS)

Wigeland, R.A.; Cahalan, J.E.

2009-01-01

Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

Human factors review for Severe Accident Sequence Analysis (SASA)

International Nuclear Information System (INIS)

Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

1984-01-01

The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure

Safety analysis for key design features of KALIMER-600 design concept

International Nuclear Information System (INIS)

Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S.

2005-03-01

KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed

Preliminary safety analysis for key design features of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

2000-07-01

KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

A task for generations. A commission plans for the future. Pt. 2. Public participation, time required, international comparison, past conflicts; Generationenaufgabe Endlagerung. Eine Kommission plant fuer die Zukunft. T. 2. Konzept zur Oeffentlichkeitsbeteiligung, Zeitbedarf, internationaler Vergleich, Konflikte der Vergangenheit

Energy Technology Data Exchange (ETDEWEB)

Fischer, Bernhard [E.ON Generation GmbH, Hannover (Germany); Jaeger, Gerd [RWTH Aachen Univ. (Germany)

2016-10-15

German Federal and State governments have committed the political foundations for the disposal of high radioactive, heat-generating waste with the Repository Site Selection Act (StandAG). The act defines a new site selection procedure and the ''Kommission Lagerung hoch radioaktiver Abfallstoffe'' (Commission Disposal of High Radioactive Waste). The Commission should evaluate the site selection process criteria, processes and decision-making basis, evaluate the StandAG and make proposals for public participation and transparency. The commission presented its final report on 5 July 2016. atw spoke with the representatives of industry, Dr. Bernhard Fischer and Prof. Dr. Gerd Jaeger, on the commission work.

A task for generations. A commission plans for the future. Pt. 1; Generationenaufgabe Endlagerung. Eine Kommission plant fuer die Zukunft. T. 1. Arbeit und Umgang in der Kommission, Entsorgungspfad, Beteiligung der Oeffentlichkeit, Entscheidungskriterien

Energy Technology Data Exchange (ETDEWEB)

Fischer, Bernhard [PreussenElektra GmbH, Hannover (Germany); Jaeger, Gerd [RWTH Aachen (Germany)

2016-08-15

German Federal and State governments have committed the political foundations for the disposal of high radioactive, heat-generating waste with the Repository Site Selection Act (StandAG). The act defines a new site selection procedure and the ''Kommission Lagerung hoch radioaktiver Abfallstoffe'' (Commission Storage of High Radioactive Waste). The Commission should evaluate the site selection process criteria, processes and decision-making basis, evaluate the StandAG and make proposals for public participation and transparency. The commission presented its final report on 5 July 2016. atw spoke with the representatives of industry, Dr. Bernhard Fischer and Prof. Dr. Gerd Jaeger, on the commission work.

U.S. advanced accelerator applications program: plans to develop and test waste transmutation technologies

International Nuclear Information System (INIS)

Van Tuyle, G.; Bennett, D.; Arthur, E.; Cappiello, M.; Finck, P.; Hill, D.; Herczeg, J.; Goldner, F.

2001-01-01

The primary mission of the U.S. Advanced Accelerator Applications (AAA) Program is to establish a national nuclear technology research capability that can demonstrate accelerator-based transmutation of waste and conduct transmutation research while at the same time providing a capability for the production of tritium if required. The AAA Program was created during fiscal year 2001 from the Accelerator Transmutation of Waste (ATW) Program and the Accelerator Production of Tritium (APT) Project. This paper describes the new AAA Program, as well as its two major components: development and testing of waste transmutation technologies and construction of an integrated accelerator-driven test facility (ADTF). (author)

Integral fast reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

DCS cabinet power loss analysis for CPR1000 nuclear power plant

International Nuclear Information System (INIS)

Zhou Liang; Zhao Yanfeng; Sun Yongbin

2014-01-01

The DCS overall structure of CRP1000 nuclear power plant was introduced. Based on the RPC, the signal interface character and signal processing mechanism on the key root were analyzed. By the power loss analyzing of RPC, the RPC loss power may lead reactor trip signal from anticipated transient without scram (ATWS) system. The results indicate that it is necessary to search DCS cabinet power loss analysis. Optimizing and assigning the main water flow signals can avoid trigger reactor trip signal by mistake. The DCS cabinet power loss analysis can optimize the I and C (instrumentation and control) design and increase the nuclear plant's reliability. (authors)

Integral Fast Reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

BWR shutdown analyzer using artificial intelligence (AI) techniques

International Nuclear Information System (INIS)

Cain, D.G.

1986-01-01

A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

Inherent controllability in modular ALMRs

International Nuclear Information System (INIS)

Sackett, J.I.; Sevy, R.H.; Wei, T.Y.C.

1989-01-01

As part of recent development efforts on advanced reactor designs ANL has proposed the IFR (Integral Fast Reactor) concept. The IFR concept is currently being applied to modular sized reactors which would be built in multiple power paks together with an integrated fuel cycle facility. It has been amply demonstrated that the concept as applied to the modular designs has significant advantages in regard to ATWS transients. Attention is now being focussed on determining whether or not those advantages deriving from the traits of the IFR can be translated to the operational/DBA (design basis accident) class of transients. 5 refs., 3 figs., 3 tabs

Nuclear power risk criteria for Mexico

International Nuclear Information System (INIS)

Meade, D.

1988-08-01

The preliminary sequence of events for three types of LOCAs (low, medium and large) and seven transients, particularly turbine trip, loss of offsite power were developed. All of the systems involved in the sequence were examined and analysed in detail and a success criterion was defined for each system in accordance with the initiating events. The difference of transient with and without SCRAM was discussed and a special sequence for the last case (ATWS) was developed. The quantification of the sequence was performed using some results from the PSA (level 1) for Laguna Verde Nuclear Power Plant (LVNPP) and the most significant sequences were shown. 16 refs, 7 figs, 7 tabs

Thorium nuclear fuel cycle technology

International Nuclear Information System (INIS)

Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

1998-03-01

Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

Preliminary safety analysis for key design features of KALIMER with breakeven core

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

2001-06-01

KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

Balance of plant modeling in TRAC-BD1/MOD1

International Nuclear Information System (INIS)

Weaver, W.L.; Giles, M.M.; Mohr, C.M.

1983-01-01

The mission of the TRAC-BD1/MOD1 code is to provide a best-estimate analysis capability for Boiling Water Reactor systems and related experimental facilities for the full range of accidents from large and small break Loss-of-Coolant accidents to operational transients including anticipated transients without scram (ATWS), for which point reactor kinetics is adequate (as a first approximation). Recent model developments allow a complete reactor system including the containment and the balance of plant to be modeled. This paper describes the balance of plant models and presents the results of a simulation of a loss-of-feedwater heater transients which was used to assess the performance of the balance of plant models

LOFA and RIA analysis of the Indonesian Multipurpose research reactor RSG-GAS 1)

International Nuclear Information System (INIS)

Endiah Puji Hastuti; Hudi Hastowo; Iman Kuntoro

1999-01-01

Investigation on accident of the Indonesian Multipurpose research reactor RSG-GAS has been performed by computer simulation technique. Two groups of transients were considered, namely transient due to loss of primary cooling system (LOFA) and power excursion due to reactivity insertion (RIA). In such a transient condition, the Common Mode Failure (CMF) is considered and it will induce a situation so called unprotected transient or Anticipated Transient Without Scram (ATWS). RELAP5, PARET-ANL and EUREKA-2RR computer packages have been applied for these analyses. Simulations result done using these computer packages showed that in the occurrence of LOFA and RIA, failure on fuel elements is limited to the region with the highest power factor. (author)

Feedback control systems for non-linear simulation of operational transients in LMFBRs

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Agrawal, A.K.; Srinivasan, E.S.

Adequate modeling of Plant Control Systems (PCS) for the study of Anticipated Transients Without Scram (ATWS) is of considerable significance in the design, operation and safety evaluation of Liquid-Metal-Cooled Fast Breeder Reactor (LMFBR) systems. To assess the system response to high frequency, low consequence events, the plant needs to be dynamically simulated. The description of analytical and numerical models for PCS that have been developed and incorporated into the loop version of the Super System Code (SSC-L) are described. The importance of detailed modeling of control systems is discussed. Sample transient results obtained for a 10% ramp change of load in 40 s in the Clinch River Breeder Reactor Plant (CRBRP) are also shown

CESEC III code conversion from Apollo to HP9000

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-01-01

CESEC code is a computer program used to analyze transient behaviour of reactor coolant systems of nuclear power plants. CESEC III is an extension of original CESEC code in order to apply wide range of accident analysis including ATWS model. Major parameters during the transients are calculated by CESEC. This report firstly describes detailed work carried out for installation of CESEC III on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also described in relation to installation of CESECIII on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

CESEC III code conversion from Apollo to HP9000

Energy Technology Data Exchange (ETDEWEB)

Lee, Hae Cho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

CESEC code is a computer program used to analyze transient behaviour of reactor coolant systems of nuclear power plants. CESEC III is an extension of original CESEC code in order to apply wide range of accident analysis including ATWS model. Major parameters during the transients are calculated by CESEC. This report firstly describes detailed work carried out for installation of CESEC III on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also described in relation to installation of CESECIII on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new.

European Nuclear Features

International Nuclear Information System (INIS)

Barre, B.; Gonzalez, E.; Diaz Diaz, J.L.; Jimenez, J.L.; Velarde, G.; Navarro, J.M.; Hittner, D.; Dominguez, M.T.; Bollini, G.; Martin, A.; Suarez, J.; Traini, E.; Lang-Lenton, J.

2004-01-01

''European Nuclear Features - ENF'' is a joint publication of the three specialized technical journals, Nuclear Espana (Spain), Revue General Nucleaire (France), and atw - International Journal of Nuclear Power (Germany). The ENF support the international Europeen exchange of information and news about energy and nuclear power. News items, comments, and scientific and technical contributions will cover important aspects of the field. The second issue of ENF contains contributions about theses topics, among others: Institutional and Political Changes in the EU. - CIEMAT Department of Nuclear Fission: A General Overview. - Inertial Fusion Energy at DENIM. - High Temperature Reactors. European Research Programme. - On Site Assistance to Khmelnitsky NPP 1 and 2 (Ukraine). - Dismantling and Decommissioning of Vandellos I. (orig.)

TRACG: Twenty years of collaboration between ENUSA and GE-HITACHI

International Nuclear Information System (INIS)

Haces, J.; Trueba, M.; Garcia, J.; Barrera, J.

2011-01-01

TRACG is the GE Hitachi Nuclear Energy (GEH) proprietary version of the Transient Reactor Analysis Code. It is a best-estimate code for analysis of boiling eater reactors (BWR). Enusa has extensively contributed to the development of TRACG, applying this code to different scenarios and BWR plants: loss-of-coolant accident (LOCA), anticipated operational occurrences (AOO), instability events licensing of GNF fuel for Nordic plants, anticipated transients without scram (ATWS) reactivity insertion accidents (RIA), validation of the simulator for the Advanced BWR (ABWR) plant, the licensing of the TRACG based U. s. Nuclear Regulatory commission (NRC)-approved AOO and LOCA licensing methodologies, and in the licensing of the passively safe generation III+ Economic Simplified Boiling Water Reactor (ESBWR).

An overview of Russian experience and capabilities for development of ATW/ABC systems

Science.gov (United States)

Kazaritsky, Vladimir D.

1995-09-01

Several Russian institutes are expected to undertake a feasibility study of nuclear power systems based on proton accelerators. The examined systems are intended for conversion of surplus Pu and transmutation of long-lived radioactive waste. This research motivated by the demilitarization agreements and criticism of traditional nuclear power is focused on environmental protection.

ATWS analyses. Analysis of anticipated transients without reactor scram in Combustion Engineering NSSS's

International Nuclear Information System (INIS)

1976-05-01

Results are presented of analyses of the transient thermal-hydraulic conditions and radiological release consequences which would occur in power plants which employ a Combustion Engineering Nuclear Steam Supply System during Anticipated Transients Without Scram due to a lack of insertion of the Control Element Assemblies upon signals for automatic or manual reactor shutdown. The transients analyzed include all events which meet the criterion to be considered as anticipated at least once in the plant lifetime with automatic reactor shutdown

An overview of Russian experience and capabilities for development of ATW/ABC systems

Energy Technology Data Exchange (ETDEWEB)

Kazaritsky, V.D. [Institute for Theoretical and Experimental Physics, Moscow (Russian Federation)

1995-10-01

Several Russian institutes are expected to undertake a feasibility study of nuclear power systems based on proton accelerators. The examined systems are intended for conversion of surplus Pu and transmutation of long-lived radioactive waste. This research motivated by the demilitarisation agreements and criticism of traditional nuclear power is focused on environmental protection.

Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Hodge, S.A.

1983-01-01

Since its inception in October 1980, the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory (ORNL) has completed four studies including Station Blackout, Scram Discharge Volume Break, Loss of Decay Heat Removal, and Loss of Injection accident sequences for the Browns Ferry Nuclear Plant. The accident analyses incorporated in a SASA study provide much greater detail than that practically achievable in a Probabilistic Risk Assessment (PRA). When applied to the candidate dominant accident sequences identified by a PRA, the detailed SASA results determine if factors neglected by the PRA would have a significant effect on the order of dominant sequences. Ongoing SASA work at ORNL involves the analysis of Anticipated Transients Without Scram (ATWS) sequences for Browns Ferry

Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

2016-08-01

This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

Fabrication of the Textured Ni-9.3at.%W Alloy Substrate for Coated Conductors

DEFF Research Database (Denmark)

Gao, M. M.; Suo, H. L.; Grivel, Jean-Claude

2011-01-01

It is difficult to obtain a sharp cube texture in the Ni-9.3at.% W substrate used for coated conductors due to its low stacking fault energy. In this paper, the traditional cold rolling procedure was optimized by introducing an intermediate recovery annealing. The deformation texture has been imp...

Analyses of Design Extended Condition Events for the Prototype Gen-IV SFR

International Nuclear Information System (INIS)

Choi, Chiwoong; Jeong, Taekyung; Lee, Kwilim; Jeong, Jaeho; Ha, Kwiseok

2015-01-01

In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). For example, the CRSS in the PRISM designed based on the 0.4 $ reactivity insertion, which is analyzed with safety analysis of UTOP event. Moreover, the sensitivity tests for weighting factor in the core radial expansion reactivity feedback model were also carried out for all ATWS events. Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W GP , which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W GP is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate expansion in the ULOHS is predominant

2010 ANNUAL MEETING ON NUCLEAR TECHNOLOGY. Pt. 3. Section reports; JAHRESTAGUNG KERNTECHNIK 2010. T. 3. Sektionsberichte Technische Sitzungen

Energy Technology Data Exchange (ETDEWEB)

Arnold, Uwe; Baumann, Erik [AREVA NP GmbH, Erlangen (Germany); Fischer, Ulrich; Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany); Gehring, Michael [Babcock Noell GmbH, Wuerzburg (Germany); Roedig, Manfred [Forschungszentrum Juelich GmbH (Germany); Willschuetz, Hans-Georg [E.ON Kernkraft GmbH, Hannover (Germany); Goers, Stefan [TUEV NORD SysTec GMbH und Co. KG, Hamburg (Germany); Schoenfelder, Christian [AREVA NP GmbH, Offenbach (Germany)

2010-12-15

Summary report on these 6 - out of 12 - Sessions of the Annual Conference on Nuclear Technology held in Berlin on May 3 to 6, 2010: - Decommissioning of Nuclear Installations (Session 7), - Fusion Technology (Session 8), - Energy Industry and Economics (Session 10), - Radiation Protection (Session 11), - New Build and Innovations (Session 12), and - Education, Expert Knowledge, Know-how-Transfer (Session 13). The other Sessions: - Reactor Physics and Methods of Calculation (Session 1), - Thermodynamics and Fluid Dynamics (Session 2), - Safety of Nuclear Installations - Methods, Analysis, Results (Session 3), - Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Session 4), - Front End of the Fuel Cycle, Fuel Elements and Core Components (Session 5), - Operation of Nuclear Installations (Session 6) have been covered in atw issues 10 and 11 (2010). (orig.)

2010 ANNUAL MEETING ON NUCLEAR TECHNOLOGY. Pt. 3. Section reports

International Nuclear Information System (INIS)

Arnold, Uwe; Baumann, Erik; Fischer, Ulrich; Bohnstedt, Angelika; Gehring, Michael; Roedig, Manfred; Willschuetz, Hans-Georg; Goers, Stefan; Schoenfelder, Christian

2010-01-01

Summary report on these 6 - out of 12 - Sessions of the Annual Conference on Nuclear Technology held in Berlin on May 3 to 6, 2010: - Decommissioning of Nuclear Installations (Session 7), - Fusion Technology (Session 8), - Energy Industry and Economics (Session 10), - Radiation Protection (Session 11), - New Build and Innovations (Session 12), and - Education, Expert Knowledge, Know-how-Transfer (Session 13). The other Sessions: - Reactor Physics and Methods of Calculation (Session 1), - Thermodynamics and Fluid Dynamics (Session 2), - Safety of Nuclear Installations - Methods, Analysis, Results (Session 3), - Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Session 4), - Front End of the Fuel Cycle, Fuel Elements and Core Components (Session 5), - Operation of Nuclear Installations (Session 6) have been covered in atw issues 10 and 11 (2010). (orig.)

Exchange of pressurizer safeguarding system at Biblis nuclear power station

International Nuclear Information System (INIS)

Weber, D.; Hofbeck, W.

1991-01-01

Valves and piping of the pressurizer safeguarding system are exchanged and reset in such a way that they are suitable not only for discharging steam, but also for discharging a water-steam mixture and hot pressurized water; for the emergency measure of primary depressurization by hand (bleed) in the event of failure of the entire feedwater supply and station black-out, and in the event of operational transients with supposed failure of the reactor scram (ATWS). To achieve this, in addition to the requirements of the pressurizer discharging station, changes have to be made to the valve drive to dominate the water loads. During the 1990 inspection this exchange of the pressurizer discharging station was performed at the Biblis A unit as the first German plant. (orig.) [de

The Plasma Window: A Windowless High Pressure-Vacuum Interface for Various Accelerator Applications

International Nuclear Information System (INIS)

Hershcovitch, A. I.; Johnson, E. D.; Lanza, R. C.

1999-01-01

The Plasma Window is a stabilized plasma arc used as an interface between accelerator vacuum and pressurized targets. There is no solid material introduced into the beam and thus it is also capable of transmitting particle beams and electromagnetic radiation with low loss and of sustaining high beam currents without damage. Measurements on a prototype system with a 3 mm diameter opening have shown that pressure differences of more than 2.5 atmospheres can be sustained with an input pressure of ∼ 10 -6 Torr. The system is capable of scaling to higher-pressure differences and larger apertures. Various plasma window applications for synchrotron light sources, high power lasers, internal targets, high current accelerators such as the HAWK, ATW, APT, DARHT, spallation sources, as well as for a number of commercial applications, is discussed

Experience with the use of programmable logic controllers in nuclear safety applications. Final report

International Nuclear Information System (INIS)

Brown, E.M.; Stofko, M.J.

1995-03-01

This report describes the implementation and experience with Programmable Logic Controllers (PLC) for nuclear safety applications. Two applications are described. The first is an Anticipated Transient Without Scram (ATWS) mitigation system provided as a Diverse Auxiliary Feedwater Actuation System (DAFAS). It was implemented at Arizona Public Service's Palo Verde Nuclear Generating Station and has been in commercial operation since early 1992. The second system described is an Emergency Diesel Generator Bus Load Sequencer installed at Florida Power and Light's Turkey Point Nuclear Power Plant. This system was installed as part of an upgrade to the emergency power system in 1988. The experience gained in the design, development, implementation and qualification of these systems will be beneficial to utilities that are considering the utilization of PLCs for their plant applications

Nuclear power 1997: Assured know-how. Session reports of the annual meeting on nuclear technology, May 13-15, 1997, in Aachen

International Nuclear Information System (INIS)

Bernnat, W.; Vayssier, G.; Unger, H.; Huettmann, A.; Sassenhagen, U.; Eickelpasch, N.; Eidens, J.

1997-01-01

After 1986 the 'Eurogress' of Aachen, was the venue of the Annual Nuclear Conference, this year on May 15, 1997. Attendance showed a slight increase, and the organizers. Deutsches Atomforum (DAtF) and the Kerntechnische Gesellschaft (KTG), welcomed more than 1000 participants at the world's biggest event of this kind. The program was arranged in the traditional proven format, with plenary sessions on the first day, and technical sessions, poster sessions, special events, and technical excursions on the other two days. These proceedings were accompanied by an exhibition arranged by vendors, suppliers, and service companies. Following the summary of the plenary day published on pp. 375 to 386 in atw 7/97, the survey in this issue covers the technical sessions as seen by the rapporteurs. (orig.) [de

Sensitivity Analysis of Uncertainty Parameter based on MARS-LMR Code on SHRT-45R of EBR II

Energy Technology Data Exchange (ETDEWEB)

Kang, Seok-Ju; Kang, Doo-Hyuk; Seo, Jae-Seung [System Engineering and Technology Co., Daejeon (Korea, Republic of); Bae, Sung-Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Hae-Yong [Sejong University, Seoul (Korea, Republic of)

2016-10-15

In order to assess the uncertainty quantification of the MARS-LMR code, the code has been improved by modifying the source code to accommodate calculation process required for uncertainty quantification. In the present study, a transient of Unprotected Loss of Flow(ULOF) is selected as typical cases of as Anticipated Transient without Scram(ATWS) which belongs to DEC category. The MARS-LMR input generation for EBR II SHRT-45R and execution works are performed by using the PAPIRUS program. The sensitivity analysis is carried out with Uncertainty Parameter of the MARS-LMR code for EBR-II SHRT-45R. Based on the results of sensitivity analysis, dominant parameters with large sensitivity to FoM are picked out. Dominant parameters selected are closely related to the development process of ULOF event.

Comparison of the SASSYS/SAS4A radial core expansion reactivity feedback model and the empirical correlation for FFTF

International Nuclear Information System (INIS)

Wigeland, R.A.

1987-01-01

The present emphasis on inherent safety for LMR designs has resulted in a need to represent the various reactivity feedback mechanisms as accurately as possible. The dominant negative reactivity feedback has been found to result from radial expansion of the core for most postulated ATWS events. For this reason, a more detailed model for calculating the reactivity feedback from radial core expansion has been recently developed for use with the SASSYS/SAS4A Code System. The purpose of this summary is to present an extension to the model so that it is more suitable for handling a core restraint design as used in FFTF, and to compare the SASSYS/SAS4A results using this model to the empirical correlation presently being used to account for radial core expansion reactivity feedback to FFTF

The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

International Nuclear Information System (INIS)

Fujita, E.K.; Wade, D.C.

1990-01-01

This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

The heating operational summarization in three winters of a 5 MW test heating reactor

International Nuclear Information System (INIS)

Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

1992-09-01

The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

Morality and ethics in high technology

International Nuclear Information System (INIS)

Schroeter, K.U.

2003-01-01

The ethical debate about what is feasible culminates, for one side, in the indignant moral question whether man is allowed to do all he is able to do and, for the other side, in the very obligation to keep redefining the limits of creation, and to act accordingly. Consequently, the Young Generation, at their meeting in Gronau, Westphalia (about which we reported), discussed about ''High Technology - Responsible on Ethical and Moral Grounds?'' The paper presented to the participants by pastor Kai Uwe Schroeter reflects this dichotomy, but also takes a clear position in favor of the expansion of nuclear power. This issue of atw contains a revised version of the paper. It is published in the hope that it will furnish arguments for the philosophical and ethical debates about high technology. (orig.) [de

Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

International Nuclear Information System (INIS)

Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

1984-01-01

This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

Safety reassessment of the Paks NPP (the AGNES project)

Energy Technology Data Exchange (ETDEWEB)

Gado, J [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; Bajsz, J; Cserhati, A; Elter, J [Paksi Atomeroemue Vallalat, Paks (Hungary); Hollo, E [Energiagazdalkodasi Intezet, Budapest (Hungary); Kovacs, K [EROTERV Engineering and Contractor Co (Hungary); Maroti, L [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; Miko, S [Paksi Atomeroemue Vallalat, Paks (Hungary); Techy, Z [Energiagazdalkodasi Intezet, Budapest (Hungary); Vidovszky, I [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics

1996-12-31

The reassessment of the Paks NPP safety according to internationally recognized criteria of the Advanced General and New Evaluation of Safety (AGNES) project is outlined. The Paks NPP consists of four WWER-440/V-213 units. The following groups of analysis have been performed: system analysis and description; analysis of design basis accidents; severe accidents analysis; level 1 probabilistic safety analysis. Postulated accidents (PA) and Anticipated Operational Occurrences (AOO) are estimated in detail for the following initiating events: increase/decrease in secondary heat removal; decrease in primary coolant inventory; increase/decrease of reactor coolant inventory; reactivity and power distribution anomalies; analysis of transients with the failure of reactor scram (ATWS); pressurized thermal shock analyses. Severe accident analysis was made for the accidents on in-vessel phase and containment phase, for radioactive release and for accident management.

On legal requirements for construction of high temperature reactors (HTR) in Poland

Energy Technology Data Exchange (ETDEWEB)

Nowacki, Tomasz R. [Ministry of Economic Development, Warsaw (Poland). Dept. for Regulatory Risk Assessment

2017-08-15

In the July 2016 issue of atw an article has been published on the legal obstacles to the construction of HTRs in Poland. The authors have raised a number of objections to the Polish law with the main thesis of the inability, or at least a significant impediment to the construction of such installations without significant legislative intervention. The main purpose of this text is to prove that the construction of HTRs based on the existing Polish laws and regulations is possible. In addition, the author intends to clarify the particular concerns expressed in the article regarding the particular legislation and correct improper statements and interpretations of the Polish nuclear law. The article deals only with strictly legal issues and does not take a stand on the technical feasibility and reality of ambitious plans for the construction of HTRs in Poland.

2010 ANNUAL MEETING ON NUCLEAR TECHNOLOGY. Pt. 4. Section reports

International Nuclear Information System (INIS)

Berlepsch, Thilo v.; Hering, Wolfgang

2011-01-01

Summary report on 2 Sessions of Section: - New Build and Innovations (Section 12) of the ANNUAL MEETING On NUCLEAR TECHNOLOGY held in Berlin on May 4 to 6, 2010. The other Sections 'Reactor Physics and Methods of Calculation (Section 1)', 'Thermodynamics and Fluid Dynamics (Section 2)', 'Safety of Nuclear Installations - Methods, Analysis, Results (Section 3)', 'Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Section 4)', 'Front End of the Fuel Cycle, Fuel Elements and Core Components (Section 5)', 'Operation of Nuclear Installations (Section 6)', 'Decommissioning of Nuclear Installations (Section 7)', 'Fusion Technology (Section 8)', 'Energy Industry and Economics (Section 10)', 'Radiation Protection (Section 11)', 'New Build and Innovations (Session New Build and Innovations, Section 12)', and 'Education, Expert Knowledge, Know-how-Transfer (Section 13)' have been covered in atw issues 10, 11 and 12 (2010). (orig.)

A study on the regulatory approach of major technical issues

Energy Technology Data Exchange (ETDEWEB)

Chang, Keun Sun; Choi, J. T.; Kim, I. J. [Sunmoon Univ., Asan (Korea, Republic of); Lee, S. H.; Baek, W. P.; Yang, S. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

1998-06-15

This project is to provide the regulatory direction of 4 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : reliability of risk significant SSCs established during design stage must be maintained through the operating life of the plant, currently used classification method of plant conditions and safety requirement were reviewed, and a quantitative classification method is needed to be developed further, the basic regulatory directions are proposed for multiple failures such as SBO, TLOFW, multiple SGTR and ATWS, safety requirements are proposed for survivability/availability of severe accident mitigation design features by 5 items if basic requirements, selection of initial event, identification of available equipment and instruments, identification of environmental conditions and verification methods.

Nuclear power 1996: potential for further development. Session reports of the annual meeting on nuclear technology, May 21-23, 1996 in Mannheim

International Nuclear Information System (INIS)

Anon.

1996-01-01

For the third time, the 'Rosengarten', the congress center of the city of Mannheim, was the venue of the Annual Nuclear Conference, this year on May 21-23, 1996. Attendance showed a slight increase, and the organizers, Deutsches Atomforum (DAtF) and the Kerntechnische Gesellschaft (KTG), welcomed more than 1000 participants at the world's biggest event of this kind. The program was arranged in the traditional, proven format, with plenary sessions on the first day, and technical sessions, poster sessions, special events, and technical excursions on the other two days. These proceedings were accompanied by an exhibition arranged by vendors, suppliers, and service companies. Following the summary of the plenary day published on pp. 385-95 in atw 6/96, the survey in this issue covers the technical sessions as seen by the rapporteurs. (orig.) [de

The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

1990-01-01

This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

GRSAC Users Manual

International Nuclear Information System (INIS)

Ball, S.J.; Nypaver, D.J.

1999-01-01

An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model

2009 annual meeting on nuclear technology. Pt. 1. Section reports; JAHRESTAGUNG KERNTECHNIK 2009. T. 1. Sektionsberichte

Energy Technology Data Exchange (ETDEWEB)

Schaffrath, Andreas [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany); Hartmann, Miks; Hoffmann, Petra Britt [Areva NP GmbH, Erlangen (Germany); Stieglitz, Robert [Forschungszentrum Karlsruhe, Eggenstein-Leopoldshafen (Germany); Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf, Dresden (Germany); Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf, Inst. fuer Sicherheitsforschung, Dresden (Germany); Hollands, Thorsten [Ruhr-Univ. Bochum (RUB), Energy Systems and Energy Economics (LEE), Bochum (Germany); Sanchez Espinoza, Victor Hugo [Forschungszentrum Karlsruhe, Inst. fuer Reaktorsicherheit, Eggenstein-Leopoldshafen (Germany); Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany); Sonnenburg, H.G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Muenchen (Germany)

2009-08-15

Summary report on these 3 - out of 13 - Sessions of the Annual Conference on Nuclear Technology held in Dresden on May 12 to 14, 2009: Thermodynamics and Fluid Dynamics (Session 2), Safety of Nuclear Installations - Methods, Analysis, Results (Session 3), and, Front End of the Fuel Cycle, Fuel Elements and Core Components (Session 4). The other Sessions Reactor Physics and Methods of Calculation (Session 1), Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Session 5), Operation of Nuclear Installations (Session 6), Decommissioning of Nuclear Installations (Session 7), Fusion Technology (Session 8), Research Reactors, Neutron Sources (Session 9), Energy Industry and Economics (Session 10), Radiation Protection (Session 11), New Build and Innovations (Session 12), and Education, Expert Knowledge, Know How Transfer (Session 13) have be covered in reports in further issues of atw. (orig.)

2009 annual meeting on nuclear technology. Pt. 1. Section reports

International Nuclear Information System (INIS)

Schaffrath, Andreas; Hartmann, Miks; Hoffmann, Petra Britt; Stieglitz, Robert; Hoehne, Thomas; Weiss, Frank-Peter; Hollands, Thorsten; Sanchez Espinoza, Victor Hugo; Tietsch, Wolfgang; Sonnenburg, H.G.

2009-01-01

Summary report on these 3 - out of 13 - Sessions of the Annual Conference on Nuclear Technology held in Dresden on May 12 to 14, 2009: Thermodynamics and Fluid Dynamics (Session 2), Safety of Nuclear Installations - Methods, Analysis, Results (Session 3), and, Front End of the Fuel Cycle, Fuel Elements and Core Components (Session 4). The other Sessions Reactor Physics and Methods of Calculation (Session 1), Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Session 5), Operation of Nuclear Installations (Session 6), Decommissioning of Nuclear Installations (Session 7), Fusion Technology (Session 8), Research Reactors, Neutron Sources (Session 9), Energy Industry and Economics (Session 10), Radiation Protection (Session 11), New Build and Innovations (Session 12), and Education, Expert Knowledge, Know How Transfer (Session 13) have be covered in reports in further issues of atw. (orig.)

Safety reassessment of the Paks NPP (the AGNES project)

International Nuclear Information System (INIS)

Gado, J.; Hollo, E.; Kovacs, K.; Maroti, L.; Techy, Z.; Vidovszky, I.

1995-01-01

The reassessment of the Paks NPP safety according to internationally recognized criteria of the Advanced General and New Evaluation of Safety (AGNES) project is outlined. The Paks NPP consists of four WWER-440/V-213 units. The following groups of analysis have been performed: system analysis and description; analysis of design basis accidents; severe accidents analysis; level 1 probabilistic safety analysis. Postulated accidents (PA) and Anticipated Operational Occurrences (AOO) are estimated in detail for the following initiating events: increase/decrease in secondary heat removal; decrease in primary coolant inventory; increase/decrease of reactor coolant inventory; reactivity and power distribution anomalies; analysis of transients with the failure of reactor scram (ATWS); pressurized thermal shock analyses. Severe accident analysis was made for the accidents on in-vessel phase and containment phase, for radioactive release and for accident management

GRSAC Users Manual

Energy Technology Data Exchange (ETDEWEB)

Ball, S.J.; Nypaver, D.J.

1999-02-01

An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model.

On legal requirements for construction of high temperature reactors (HTR) in Poland

International Nuclear Information System (INIS)

Nowacki, Tomasz R.

2017-01-01

In the July 2016 issue of atw an article has been published on the legal obstacles to the construction of HTRs in Poland. The authors have raised a number of objections to the Polish law with the main thesis of the inability, or at least a significant impediment to the construction of such installations without significant legislative intervention. The main purpose of this text is to prove that the construction of HTRs based on the existing Polish laws and regulations is possible. In addition, the author intends to clarify the particular concerns expressed in the article regarding the particular legislation and correct improper statements and interpretations of the Polish nuclear law. The article deals only with strictly legal issues and does not take a stand on the technical feasibility and reality of ambitious plans for the construction of HTRs in Poland.

Accelerator driven reactors and nuclear waste management projects in the Czech Republic

Energy Technology Data Exchange (ETDEWEB)

Janouch, F. [Royal Institute of Technology, Stockholm (Sweden); Mach, R. [Institute of Nuclear Physics, Rez near Prague (Czechoslovakia)

1995-10-01

The Czech Republic is almost the only country in the central Europe which continues with the construction of nuclear power reactors. Its small territory and dense population causes public worries concerning the disposal of the spent nuclear fuel. The Czech nuclear scientists and the power companies and the nuclear industries are therefore looking for alternative solutions. The Los Alamos ATW project had received a positive response in the Czech mass-media and even in the industrial and governmental quarters. The recent scientific symposium {open_quotes}Accelerator driven reactors and nuclear waste management{close_quotes} convened at the Liblice castle near Prague, 27-29. 6. 1994 and sponsored by the Czech Energy Company CEZ, reviewed the competencies and experimental basis in the Czech republic and made the first attempt to formulate the national approach and to establish international collaboration in this area.

Sensitivity Tests for the Unprotected Events of the Prototype Gen-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Choi, Chiwoong; Lee, Kwilim; Jeong, Jaeho; Yu, Jin; An, Sangjun; Lee, Seung Won; Chang, Wonpyo; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Unprotected Transient Over Power, (UTOP), Unprotected Loss Of Flow (ULOF), and Unprotected Loss Of Heat Sink (ULOHS) are selected as ATWS events. Among these accidents, the ULOF event shows the lowest clad temperature. However, the ULOHS event showed the highest peak clad temperature, due to the positive CRDL/RV expansion reactivity feedback and insufficient DHRS capacity. In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate

Overview of NSSS Fluid System Design of PGSFR

Energy Technology Data Exchange (ETDEWEB)

Han, Ji-Woong; Choi, Seok-Ki; Kim, Seong-O; Kim, Eui-Kwang; Kim, Dehee; Hong, Jonggan; Ye, Huee-Youl; Yeom, Sujin; Ryu, Seungho; Yoon, Jung; Choi, Sun Rock; Park, Jin-Seok; Lee, Tae-Ho Lee [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this paper an overview on the NSSS fluid system design of PGSFR is described based on the issued design documents. System concepts and major components design concepts for PHTS, IHTS, DHRS and SWRPRS were developed. Thermal-hydraulic characteristics were analyzed based on CFD simulation. The design bases and concepts for auxiliary systems were also developed. The upstream design requirements of fluid system such as system design requirements, component design requirements, I and C design requirements, BOP interface design requirements, design guides and P and IDs were produced. The control logic and computer code for the analysis for operational characteristics is under progress. The protection system consists of a safety grade PPS and a non-safety grade DPS (Diverse Protection System). The DPS provides a diverse method to trip the reactor to satisfy the requirements relative to ATWS (Anticipated Transients Without Scram) as well as Defense-In-Depth and Diversity.

Activities report of Instituto de Engenharia Nuclear - 1993

International Nuclear Information System (INIS)

1994-04-01

This is an activities report of 1993 year from Nuclear Engineering Institute - IEN, Brazilian CNEN. It achieves research and development in science and technology in nuclear areas and correlative, generating knowledge, products and services in society benefit. Its major activities are: production and supply of 30 equipment for radioprotection and nuclear physics, and the project and construction of monitoring system of neutron flux and radioprotection of Triga reactor of CDTN; project and construction of an actuation/attenuation circuit system of ATWS (anticipated transient without scram) at Angra-1 nuclear power plant; obtention of rare earths; development and optimization of a computational program to three-dimensional fluid mechanic - NSD 3D; study of technical and economical feasibility to the implantation, at IEN, of a big size irradiator for the sterilization, using ionizing radiation, of hospital, pharmaceutical and cosmetic goods. (C.M.). 7 tabs, 2 graphs

ISP Descriptions of Four Military Computer Architectures

Science.gov (United States)

1978-04-01

SOn Tot) ENDI IENC JEP JOPi. I JUMP ON 000 PARITY BE GIN OP TD I I GLTO NEXT T7A - ’rM AND TAI NEXT PARITY •TEXT I CEtCk PARITY (TI TO .> P * so Tot...WIo10lc6iO1 I NEXT NUIAO I STOPE (T019) IN PCcWI&ATW? (As) I NEXT EN~i ILNO HSC? UvP 7,1!sr Iwx7inr-non 1.tt-1 I tD 3-7,b-?S 11806 wiCI ~utJ-I8O Page 91...8217 ((C t4,3 YPMiNPE~ SF * d.ttýp(iiIZ) L53 MAWR.u, 1, NO I COGl. - IPIOI GC..’r ".Ced half r"o.’iv - I P411 11.Pi-4i * M,’r~pCINUCS, P’PIOWfu-.",,I)Q2𔃺i

Concept of the new generation high safety liquid metal reactor (LMFR)

International Nuclear Information System (INIS)

Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

1988-01-01

The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

Study on diagnostic plant analyzer method for support of emergency operation

International Nuclear Information System (INIS)

Yoshikawa, H.; Gofuku, A.; Itoh, K.; Wakabayashi, J.

1986-01-01

Methods of time-critical diagnostic plant analyzer are investigated which would serve as support to emergency operation of nuclear power plant. A faster-than-real-time simulator code TOKRAC is developed for analyzing PWR primary loop thermo-hydraulics of small-break LOCA and it is applied for a numerical experiment of initial phase of TMI-2 accident. TOKRAC resulted in a good agreement with a RELAP4/MOD6 calculation and the plant record with as fast as can one-tenth of real time. A real-time estimator of SG heat transfer rate based on Kalman filter is proposed and its applicability is verified using LOFT ATWS experimental data. With regards to how to integrate those methods into the software system in operation support center, a new concept of module-based simulation system is proposed which aims at offering a flexible and human-cognitive oriented environment for various analytical tool development

RAMONA-3B/MINET composite representation of BWR thermal-hydraulic systems

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.; Cazzoli, E.G.; Nepsee, T.C.; Guppy, J.G.

1985-01-01

The modification and interfacing of two computer codes, RAMONA-3B and MINET, for the thermal hydraulic transient analysis of a Boiling Water Reactor nuclear steam supply system, is described. The RAMONA-3B code provides for multi-channel thermal hydraulics and three-dimensional (or one-dimensional) neutron kinetics analysis of a boiling water reactor core. The RAMONA-3B system representation terminates at the end of the steam line and at the junction of the feedwater line at the vessel inlet. By interfacing RAMONA-3B with MINET, a generic balance-of-plant systems analysis code, a complete BWR systems code with detailed core modeling was obtained. The result is a code of particular importance to the analysis of transients such as ATWS. A comparison between the 3-D and 1-D neutronics representation is provided, along with a test case utilizing the composite RAMONA-3B/MINET code

78 FR 5864 - Waiver of Aeronautical Land-Use Assurance: Outagamie County Regional Airport (ATW), Appleton, WI

Science.gov (United States)

2013-01-28

... operation of a Public Safety Training Center (PSTC) by the Fox Valley Technical College (FVTC). The PSTC is an educational campus intended to provide degree/diploma/certificate programs to students enrolled in...-55-0002-30-06. A fair market value (FMV) appraisal for the parcel was completed in 2011 in accordance...

Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

International Nuclear Information System (INIS)

Song, Yo Taik

2001-01-01

Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented

Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

Energy Technology Data Exchange (ETDEWEB)

Song, Yo Taik

2001-01-01

Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented.

Control Rod Withdrawal Events Analyses for the Prototype Gen-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Choi, Chiwoong; Ha, Kwiseo; Jeong, Taekyeong; Jeong, Jaeho; Chang, Wonpyo; Lee, Seungwon; An, Sangjun; Lee, Kwilim [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

To confirm the limiting condition, based on the maximum allowable reactivity insertion of 0.3 $, three cases from the end of cycle (EOC) are selected. In addition, assuming the failure of CRSS by earthquake, additional cases is defined at beginning of cycle (BOC). When the CRW occurs, the reactor can be protected by plant protection system (PPS). In this study, PPS mechanism is sequentially studied for all initiating events. For design basis accidents (DBA), the reactor can be scrammed by reactor protection system (RPS). The first and seconds RPS signals are checked during transients. When RPS is failed, so called as anticipated transient without scram (ATWS), the reactor will be protected by diverse protection system (DPS). In this study, in order to analyze various initiating events related control rod withdrawal, four kinds of operating condition is defined. TOP events are analyzed using MARS-LMR. The influence of various plant protection system such as RPS and DPS are investigated.

Modeling a nuclear reactor for experimental purposes

International Nuclear Information System (INIS)

Berta, V.T.

1980-01-01

The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended functions during occurrences of off-normal conditions in a commercial nuclear reactor. Off-normal conditions arising from large and small break loss-of-coolant accidents (LOCA), operational transients, and anticipated transients without scram (ATWS) were to be investigated. This paper describes the LOFT model of the generic PWR and summarizes the experiments that have been conducted in the context of the significant findings involving the complex transient thermal-hydraulics and the consequent effects on the commercial reactor analytical licensing techniques. Through these techniques the validity of the LOFT model as a scaled counterpart of the generic PWR is shown

Response of actinides to flux changes in high-flux systems

International Nuclear Information System (INIS)

Sailor, W.C.

1993-01-01

When discussing the transmutation of actinides in accelerator-based transmutation of waste (ATW) systems, there has been some concern about the dynamics of the actinides under high transient fluxes. For a pure neptunium feed, it has been estimated that the 238 Np/ 237 Np ratio increase due to an increasing flux may lead to an unstable, positive reactivity growth. In this analysis, a perturbation method is used to calculate the response of the entire set of actinides in a general way that allows for more species than just neptunium. The time response of the system can be calculated; i.e., a plot of fuel composition and reactivity versus time after a change in flux can be made. The effects of fission products can also be included. The procedure is extremely accurate on short time scales (∼ 1000 s) for the flux levels we contemplate. Calculational results indicate that the reactivity insertions are always smaller than previously estimated

Review of the Technical Status on the Debris Bed Cooling Model

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

2007-09-15

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.

Review of the Technical Status on the Debris Bed Cooling Model

International Nuclear Information System (INIS)

Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

2007-09-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris

Safety performance of preliminary KALIMER conceptual design

Energy Technology Data Exchange (ETDEWEB)

Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

1999-07-01

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

Safety performance of preliminary KALIMER conceptual design

International Nuclear Information System (INIS)

Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

1999-01-01

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

Transmutation of fission products and actinide waste at Hanford

Energy Technology Data Exchange (ETDEWEB)

Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

1995-10-01

The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

Analysis of regulatory requirement for beyond design basis events of SMART

International Nuclear Information System (INIS)

Kim, W. S.; Seol, K. W.

2000-01-01

To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

Incentive regulation - an abstract term coming to life

International Nuclear Information System (INIS)

Heller, W.

2006-01-01

Well over one year ago, the new German Power Economy Act entered into force after prolonged debates and recurrent delays in the course of parliamentary discussions and after a mediation procedure. Essentially, the new legislation is about adopting strict rules for the networks for transport and distribution of electricity and gas, which are deemed to be natural monopolies, but also for power supply in general, and having these rules monitored and advanced by a new federal regulatory authority. This marks the transformation into national law of the so-called EU Acceleration Directives for Completion of the Single Market for Electricity and Gas of August 2003 (cf. atw 5/2005, p. 378). Although this transformation of the Acceleration Directives happened with one year's delay, it seems to have been successful in Germany, according to an assessment by the EU Commission. Other member countries had to be shown the red card by the Commission, and proceedings for violation of the Treaty had to be instituted against Spain. (orig.)

The NUREG-1150 probabilistic risk assessment for the Grand Gulf nuclear station

International Nuclear Information System (INIS)

Brown, T.D.; Breeding, R.J.; Jow, H.N.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.

1992-01-01

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the 'back-end' analyses, that is, the acident progression, source term, consequence analsyes, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term blackout group and the anticiptated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term. (orig.)

Development of Adiabatic Doppler Feedback Model in 3D space time analysis Code ARCH

International Nuclear Information System (INIS)

Dwivedi, D.K.; Gupta, Anurag

2015-01-01

Integrated 3D space-time neutron kinetics with thermal-hydraulic feedback code system is being developed for transient analysis of Compact High Temperature Reactor (CHTR) and Advanced Heavy Water Reactor (AHWR). ARCH (code for Analysis of Reactor transients in Cartesian and Hexagon geometries) has been developed with IQS module for efficient 3D space time analysis. Recently, an adiabatic Doppler (fuel temperature) feedback module has been incorporated in this ARCH-IQS version of tile code. In the adiabatic model of fuel temperature feedback, the transfer of the excess heat from the fuel to the coolant during transient is neglected. The viability of Doppler feedback in ARCH-IQS with adiabatic heating has been checked with AER benchmark (Dyn002). Analyses of anticipated transient without scram (ATWS) case in CHTR as well as in AHWR have been performed with adiabatic fuel temperature feedback. The methodology and results have been presented in this paper. (author)

Design concept of KALIMER-600

International Nuclear Information System (INIS)

Hahn, Dohee; Kim, Yeong-Il; Kim, Seong-O; Lee, Jae-Han; Lee, Yong-Bum

2005-01-01

KALIMER-600 is a pool-type sodium-cooled reactor loaded with U-TRU-10%Zr metal fuels generating the net electricity output of 600 MWe. In order to enhance the proliferation resistance, no blanket assemblies are loaded in the core. To suppress the high power peaking factor, some of the fuel rods are replaced with B 4 C rods and dummy rods. The heat transport system is comprised of two independent loops of IHTS and SGS and the safety-grade residual heat removal system, PDRC, is a completely passive system. Main features of the mechanical structure design of KALIMER-600 are the seismically isolated reactor building, the reduced total pipe length of the IHTS, the simplified reactor support, and the compact reactor internal structures. From the safety analyses, the KALIMER-600 design is verified to be capable of accommodating all the analyzed ATWS events. This self-regulation capability of the KALIMER-600 is mainly due to the inherent reactivity feedback mechanisms and completely passive PDRC system. (author)

Probabilistic risk assessment course documentation. Volume 1: PRA fundamentals

International Nuclear Information System (INIS)

Breeding, R.J.; Leahy, T.J.; Young, J.

1985-08-01

The full range of PRA topics is presented, with a special emphasis on systems analysis and PRA applications. Systems analysis topics include system modeling such as fault tree and event tree construction, failure rate data, and human Reliability. The discussion of PRA applications is centered on past and present PRA based programs, such as WASH-1400 and the Interim Reliability Evaluation Program, as well as on some of the potential future applications of PRA. The relationship of PRA to generic safety issues such as station blackout and Anticipated Transient Without Scram (ATWS) is also discussed. In addition to system modeling, the major PRA tasks of accident process analysis, and consequence analysis are presented. An explanation of the results of these activities, and the techniques by which these results are derived, forms the basis for a discussion of these topics. An additional topic which is presented in this course is the topic of PRA management, organization, and evaluation. 84 figs., 41 tabs

The German act on the reorganisation of responsibility in nuclear waste management; Des Gesetz zur Neuordnung der Verantwortung in der kerntechnischen Entsorgung

Energy Technology Data Exchange (ETDEWEB)

Raetzke, Christian [CONLAR Consulting on Nuclear Law and Regulation, Leipzig (Germany)

2017-04-15

The author discussed the Draft on the Act in the Reorganisation of Responsibility in Nuclear Waste Management in atw 12 (2016). Now, amendments are discussed, which resulted from the legislative procedure until today's draft. Significant additions affect the authorisation for the conclusion of a public-law contract between the Federal Government and the nuclear power plant operators, the deadline for the payment of the basic amount, and the option for the operation of the interim storage facilities for a transitional period by the operators on behalf of the federal company. Since the adoption of the draft act, it has become clear that the nuclear power plant operators will pay the risk premium. This will fulfil the full logic of the new system. It has also become known, that the public law contract is now ready for signing. According to the author, the act will bring a final arrangement for financing nuclear waste disposal. However, adjustment can not be avoided in practice. The concrete implementation will be a exciting topic in many ways.

Development of Risk Management Technology/Development of Risk-Informed Application Technology

Energy Technology Data Exchange (ETDEWEB)

Yang, Joon Eon; Kim, K. Y.; Ahn, K. I.; Lee, Y. H.; Lim, H. G.; Jung, W. S.; Choi, S. Y.; Han, S. J.; Ha, J. J.; Hwang, M. J.; Park, S. Y.; Yoon, C

2007-06-15

This project aims at developing risk-informed application technologies to enhance the safety and economy of nuclear power plant altogether. For this, the Integrated Level 1 and 2 PSA model is developed. In addition, the fire and internal flooding PSA models are improved according to the PSA standard of U.S.A. To solve the issues of domestic PSA model, the best-estimate thermal hydraulic analyses are preformed for the ATWS and LSSB. In order to reduce the uncertainty of PSA, several new PSA technologies are developed: (1) more exact quantification of large fault tree, (2) importance measure including the effects of external PSA. As feasibility studies of Option 2 and 3, the class of 6 systems' SSC are re-classified based on the risk information and the sensitivity analyses is performed for the EDG starting time, respectively. It is also improved that the methodology to identify the vital area of NPP. The research results of this project can be used in the regulatory body and the industry projects for risk-informed applications.

Simulation of the chemical state of irradiated oxide fuel; influence of the internal corrosion on the mechanical properties of Zry-4 tubing

International Nuclear Information System (INIS)

Hofmann, P.

1979-03-01

Zircaloy is not compatible with oxide fuel nor with some fission product elements. Therefore, chemical interaction between the irradiated oxide fuel and the Zry cladding material take place, especially at temperatures that can be reached during reactor incidents (ATWS, LOCA). In order to find out which influence the chemical interaction between the fission products and the Zry cladding material have on the mechanical properties of Zry-4 tubing out-of-pile burst experiments and creep rupture tests have been performed at temperatures >=600 0 C with short tube specimens containing simulated fission products. First of all, assessments of the chemical state of irradiated oxide fuel were performed and a method is described for introducing simulated fission product species into fresh oxide fuel for irradiation tests. As the test results of the out-of-pile studies show, only iodine can lead to a low ductility failure of the Zry-tubing at temperatures >=600 0 C. However, the influence of iodine on the deformation behavior of Zry-tubing can be neglected above 850 0 C. (orig.) [de

ESBWR power maneuvering via feedwater temperature control

International Nuclear Information System (INIS)

Saha, P.; Marquino, W.; Tucker, L. J.

2008-01-01

The ESBWR is a Generation III+ Boiling Water Reactor (BWR) driven by natural circulation. For a given geometry/hardware, system pressure, downcomer water level and feedwater temperature, the core flow rate in the ESBWR is only a function of reactor power, controlled through the control blade movement. In order to provide operational flexibility, another method of core-wide or global power maneuvering via feedwater temperature control has been developed. This is independent of power maneuvering via control blade movement, and it lowers the linear heat generation rate (LHGR) changes near the tip of control blades, which improves fuel reliability. All required stability, anticipated operational occurrences (AOOs), infrequent events, special events including anticipated transients without scram (ATWS), and loss-of-coolant accident (LOCA) analyses have been performed for the 4500 MWt ESBWR. Based on the results of these analyses at 'high', nominal and 'low' feedwater temperatures, a safe Power - Feedwater Temperature operating domain has been developed. This paper summarizes the results of these analyses and presents the ESBWR Power - Feedwater Temperature operating domain or map. (authors)

No consensus for consensus's sake. Interview with DAtF-President Dr. Otto Majewski

International Nuclear Information System (INIS)

Anon.

1999-01-01

The German power industry is not prepared to make special sacrifices at the expense of its shareholders, staff and customers merely to cure the apparent internal problems of the German government coalition and help the Greens to regain lost votes. Unless the Federal Minister for the Environment rescinds his current cascade of conditions and requirements impeding the efforts on the part of operators to resume fuel transports, everything 'boils down to a court case'. That the French government, contrary to general expectations, had not given the green light this year for the construction of a first EPR plant was due primarily to internal political reasons, which had to be respected. It did not really matter whether that decision was taken in 2000 or 2001. The important thing was that the EPR was built at all, which was beyond any doubt. This is the gist of the statements by Dr. Otto Majewski, President of the German Atomic Forum and Chief Executive Officer of Bayernwerk AG, in the interview granted atw Editor-in-Chief Gerhard Kuebler and printed in this issue of 'Atomwirtschaft'. (orig.) [de

Transient performance of S-prism

International Nuclear Information System (INIS)

Dubberley, A.E.; Boardman, C.E.; Gamble, R.E.; Hiu, M.M.; Lipps, A.J.; Wu, T.

2001-01-01

S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. Based on the success of the previous DOE sponsored Advanced Liquid Metal Reactor (ALMR) program GE has continued to develop and assess the technical viability and economic potential of an up-rated plant called SuperPRISM (S-PRISM). This paper presents the results of transient analyses performed to assess the ability of S-PRISM to accommodate severe accident initiator events. A unique safety capability of S-PRISM is accommodation of the ''higher probability'' accident initiators that led to core melt accidents in prior large LMRs. These events, the Anticipated Transients Without Scram (ATWS) events, are thus the focus of passive safety confirmation analyses. The events included in this assessment are: Unprotected Loss of Flow, Unprotected Loss of Heat Sink, Unprotected Loss of Flow and Heat sink, Unprotected Transient Overpower and Unprotected Safe Shutdown Earthquake. (author)

Development of Risk Management Technology/Development of Risk-Informed Application Technology

International Nuclear Information System (INIS)

Yang, Joon Eon; Kim, K. Y.; Ahn, K. I.; Lee, Y. H.; Lim, H. G.; Jung, W. S.; Choi, S. Y.; Han, S. J.; Ha, J. J.; Hwang, M. J.; Park, S. Y.; Yoon, C.

2007-06-01

This project aims at developing risk-informed application technologies to enhance the safety and economy of nuclear power plant altogether. For this, the Integrated Level 1 and 2 PSA model is developed. In addition, the fire and internal flooding PSA models are improved according to the PSA standard of U.S.A. To solve the issues of domestic PSA model, the best-estimate thermal hydraulic analyses are preformed for the ATWS and LSSB. In order to reduce the uncertainty of PSA, several new PSA technologies are developed: (1) more exact quantification of large fault tree, (2) importance measure including the effects of external PSA. As feasibility studies of Option 2 and 3, the class of 6 systems' SSC are re-classified based on the risk information and the sensitivity analyses is performed for the EDG starting time, respectively. It is also improved that the methodology to identify the vital area of NPP. The research results of this project can be used in the regulatory body and the industry projects for risk-informed applications

Development of nuclear transmutation technology - A study on accelerator-driven transmutation of long-lived radionuclide

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; Chung, Kie Hyung; Hong, Sang Hee; Hwang, Il Soon; Park, Byung Gi; Yang, Hyung Lyeol; Kim, Duk Kyu; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

1996-07-01

The objective of this study is to help establish the long-range nuclear waste disposal strategy through the investigations and comparisons of various= concepts of the accelerator-driven nuclear waste transmutation reactors, which have been suggested to replace the geological waste disposal due to the technical uncertainties in the long-time scale. Nuclear data, categorized in high -and low-energy neutron cross-sections, were investigated and the structures, principles, and recent progresses of proton linac were reviews, Also the accelerator power for transmutation and the economics were referred, The comparison of the transmutation concepts concentrated on two: Japanese OMEGA program of alloy fuelled system, Minor actinide molten salt system, and Eutectic alloy system and American ATW program of aqueous system and molten salt system. From the comparative study, a state-of-art of the technology has been identified as a concept employing proton-accelerate of 800 {approx} 1600 MeV with 100 mA capacity combined with liquid lead target, molten salt blanket and on-line chemical separation using centrifuge and electrowinning technology. 34 refs., 25 tabs., 64 figs. (author)

Chemically deposed layer sytems for the realization of YBa2Cu3O7-δ band conductors

International Nuclear Information System (INIS)

Engel, Sebastian

2009-01-01

The aim of this thesis was to produce new buffer-layer systems for biaxially texturated Ni5at%W substrates by means of chemical processes. As very promising materials for the buffer layers CaTiO 3 and SrTiO 3 were chosen. The production of the single layers pursued from the organometallic prestage by means of dip coating and subsequent head treatment. During the work first the single precursor solutions were to be developed. A main component of the theses forms the understanding of the texture development during the heat treatment of precursor layers on biaxially texturated metallic substrates. Based on this the growth of thick buffer layers is studied and by means of YBCO layers, which were deposed by beans of a pulsed laser, the functionality of the synthesized buffer layers proved. A further component of this thesis formes the influence of nanoscaling precipitations in thew YBCO on its superconducting properties. The YBCO deposition pursued via a variation of the TFA process, as substrate (001)-oriented SrTiO 3 monocrystals were applied

Distributed Aerodynamic Sensing and Processing Toolbox

Science.gov (United States)

Brenner, Martin; Jutte, Christine; Mangalam, Arun

2011-01-01

A Distributed Aerodynamic Sensing and Processing (DASP) toolbox was designed and fabricated for flight test applications with an Aerostructures Test Wing (ATW) mounted under the fuselage of an F-15B on the Flight Test Fixture (FTF). DASP monitors and processes the aerodynamics with the structural dynamics using nonintrusive, surface-mounted, hot-film sensing. This aerodynamic measurement tool benefits programs devoted to static/dynamic load alleviation, body freedom flutter suppression, buffet control, improvement of aerodynamic efficiency through cruise control, supersonic wave drag reduction through shock control, etc. This DASP toolbox measures local and global unsteady aerodynamic load distribution with distributed sensing. It determines correlation between aerodynamic observables (aero forces) and structural dynamics, and allows control authority increase through aeroelastic shaping and active flow control. It offers improvements in flutter suppression and, in particular, body freedom flutter suppression, as well as aerodynamic performance of wings for increased range/endurance of manned/ unmanned flight vehicles. Other improvements include inlet performance with closed-loop active flow control, and development and validation of advanced analytical and computational tools for unsteady aerodynamics.

Selection of flowing liquid lead target structural materials for accelerator driven transmutation applications

International Nuclear Information System (INIS)

Park, J.J.; Buksa, J.J.

1994-01-01

The beam entry window and container for a liquid lead spallation target will be exposed to high fluxes of protons and neutrons that are both higher in magnitude and energy than have been experienced in proton accelerators and fission reactors, as well as in a corrosive environment. The structural material of the target should have a good compatibility with liquid lead, a sufficient mechanical strength at elevated temperatures, a good performance under an intense irradiation environment, and a low neutron absorption cross section; these factors have been used to rank the applicability of a wide range of materials for structural containment Nb-1Zr has been selected for use as the structural container for the LANL ABC/ATW molten lead target. Corrosion and mass transfer behavior for various candidate structural materials in liquid lead are reviewed, together with the beneficial effects of inhibitors and various coatings to protect substrate against liquid lead corrosion. Mechanical properties of some candidate materials at elevated temperatures and the property changes resulting from 800 MeV proton irradiation are also reviewed

Thermal neutron cross section measurements for technetium-99

International Nuclear Information System (INIS)

Yates, M.A.; Schroeder, N.C.; Fowler, M.M.

1993-01-01

Technetium, because of its long half-like (213,000 years) and ability to migrate in the environment, is a primary contributor to the long-term radioactivity related risk associated with geologic nuclear waste disposal. One proposal for converting technetium to an environmentally benign element investigating transmutation with an accelerator-based system, (i.e., Accelerator Transmutation of Waste, ATW). Planning for efficient processing of technetium through the transmuter will require knowledge of the thermal neutron cross section for the 99 Tc (n,γ) 100 Tc reaction. The authors have recently remeasured this cross section. Weighed aliquots (19-205 μg) of a NIST traceable 99 Tc standard were irradiated for 30-150 sec using the pneumatic open-quotes rabbitclose quotes system of LANL's Omega West Reactor. The two gamma rays from the 15.7-sec half-life product were measured immediately after irradiation on a high-resolution Ge detector. Thermal fluxes were measured using gold foils and Cd wrapped gold foils. The observation cross section is 19 ± 1 b. This agrees well with the 1977 value but has half the uncertainty

Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

The German act on the reorganisation of responsibility in nuclear waste management

International Nuclear Information System (INIS)

Raetzke, Christian

2017-01-01

The author discussed the Draft on the Act in the Reorganisation of Responsibility in Nuclear Waste Management in atw 12 (2016). Now, amendments are discussed, which resulted from the legislative procedure until today's draft. Significant additions affect the authorisation for the conclusion of a public-law contract between the Federal Government and the nuclear power plant operators, the deadline for the payment of the basic amount, and the option for the operation of the interim storage facilities for a transitional period by the operators on behalf of the federal company. Since the adoption of the draft act, it has become clear that the nuclear power plant operators will pay the risk premium. This will fulfil the full logic of the new system. It has also become known, that the public law contract is now ready for signing. According to the author, the act will bring a final arrangement for financing nuclear waste disposal. However, adjustment can not be avoided in practice. The concrete implementation will be a exciting topic in many ways.

Chemically deposed layer sytems for the realization of YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} band conductors; Chemisch deponierte Schichtsysteme zur Realisierung von YBa{sub 2}Cu{sub 3}O{sub 7-{delta}}-Bandleitern

Energy Technology Data Exchange (ETDEWEB)

Engel, Sebastian

2009-04-30

The aim of this thesis was to produce new buffer-layer systems for biaxially texturated Ni5at%W substrates by means of chemical processes. As very promising materials for the buffer layers CaTiO{sub 3} and SrTiO{sub 3} were chosen. The production of the single layers pursued from the organometallic prestage by means of dip coating and subsequent head treatment. During the work first the single precursor solutions were to be developed. A main component of the theses forms the understanding of the texture development during the heat treatment of precursor layers on biaxially texturated metallic substrates. Based on this the growth of thick buffer layers is studied and by means of YBCO layers, which were deposed by beans of a pulsed laser, the functionality of the synthesized buffer layers proved. A further component of this thesis formes the influence of nanoscaling precipitations in thew YBCO on its superconducting properties. The YBCO deposition pursued via a variation of the TFA process, as substrate (001)-oriented SrTiO{sub 3} monocrystals were applied.

Risk Assessment Review Group report to the U.S. Nuclear Regulatory Commission

International Nuclear Information System (INIS)

Lewis, H.W.; Budnitz, R.J.; Kouts, H.J.C.; Loewenstein, W.B.; Rowe, W.D.; von Hippel, F.; Zachariasen, F.

1978-09-01

The Risk Assessment Review Group was organized by the U.S. Nuclear Regulatory Commission on July 1, 1977, with four elements to its charter: clarify the achievements and limitations of WASH-1400, the ''Rasmussen Report''; assess the peer comments thereon, and responses to those comments; study the present state of such risk assessment methodology; and recommend to the Commission how (and whether) such methodology can be used in the regulatory and licensing process. Areas of study include: risk assessment methodologies; statistical issues; completeness; the data base; and the WASH-1400 assessment of the damage to human health from radiation after a postulated accident. Specific items discussed include: Browns Ferry; common cause failure; human factors; format and scrutability; the peer review process; earthquakes; risk perception; allegations by UCS concerning WASH-1400 treatment of quality assurance and quality control; current role of probabilistic methods in the regulatory process; acts of violence; ATWS; influence of design defects in quality assurance failures; and calculation of population doses from given releases of radionuclides

The mechanism of the nano-CeO2 films deposition by electrochemistry method as coated conductor buffer layers

International Nuclear Information System (INIS)

Lu, Yuming; Cai, Shuang; Liang, Ying; Bai, Chuanyi; Liu, Zhiyong; Guo, Yanqun; Cai, Chuanbing

2015-01-01

Highlights: • Crack-free CeO 2 film thicker than 200 nm was prepared on NiW substrate by ED method. • Different electrochemical processes as hydroxide/metal mechanisms were identified. • The CeO 2 precursor films deposited by ED method were in nano-scales. - Abstract: Comparing with conventional physical vapor deposition methods, electrochemistry deposition technique shows a crack suppression effect by which the thickness of CeO 2 films on Ni–5 at.%W substrate can reach a high value up to 200 nm without any cracks, make it a potential single buffer layer for coated conductor. In the present work, the processes of CeO 2 film deposited by electrochemistry method are detailed investigated. A hydroxide reactive mechanism and an oxide reactive mechanism are distinguished for dimethyl sulfoxide and aqueous solution, respectively. Before heat treatment to achieve the required bi-axial texture performance of buffer layers, the precursor CeO 2 films are identified in nanometer scales. The crack suppression for electrochemistry deposited CeO 2 films is believed to be attributed to the nano-effects of the precursors

On the safety of French nuclear power plants. Zur Sicherheit der franzoesischen Kernkraftwerke

Energy Technology Data Exchange (ETDEWEB)

Anon,

1990-04-01

An allegedly secret report by the inspector general for nuclear safety, of EDF, has recently been unearthed and published by the French weekly 'Le Canard Enchaine, and the response in France, and very soon after also in West Germany, has been a number of alarming reports and articles in the press. Readers in West Germany have been stirred up by press reports that made French nuclear power plants appear to be a herd of hazards, which of course again added fuel to the feeling of fear of nuclear power already existing in the population. A copy of the internal report in question was sent without any fuss upon request by the atw editorial office who was preparing the interview. The report is a sober account of the state and operating behaviour of French nuclear power plants, also stating weak points seen by the safety expert that need particular attention. Materials are a main aspect in this context, particularly the materials behaviour in steam raising units. The problems have been spotted, and are given due attention. (orig./HP).

Secondary production at the Polar Front, Barents Sea, August 2007

Science.gov (United States)

Basedow, Sünnje L.; Zhou, Meng; Tande, Kurt S.

2014-02-01

To investigate spatial patterns of secondary production we sampled four core hydrographical regions of the Polar Front in the Barents Sea (Arctic Water, ArW; Polar Front Water, PFW; Atlantic Water, AtW; and Melt Water, MW) by towing an undulating instrument platform along a transect crossing the front from August 8-9, 2007. Sensors mounted on the platform provided data on the hydrography (CTD), fluorescence (Fluorometer, F) and zooplankton abundance in the size range between 0.1 and 30 mm (Laser Optical Plankton Counter, LOPC). These continuous, biophysical data with high-spatial resolution were supplemented by discrete water and zooplankton net samples at stations for sensor calibrations. After in depth quality assessments of the biophysical data, estimates were made of the vital rates based on biovolume spectrum theory. Five size groups were distinguished from the LOPC data: small (S), mainly Oithona spp. and the appendicularian Fritillaria sp.; medium (M), mainly Pseudocalanus spp. and Calanus spp. CI-CIII; large (L), mainly Calanus spp. CIV-CV; and extra large (XL and 2XL), juvenile and adult euphausids. Size groups were further divided based on transparency of organisms. Vital rates based on the biophysical in situ data in combination with biovolume spectrum theories agreed generally well with data from empirical and numerical models in the literature. ArW was characterised by subsurface maxima of chlorophyll a (chl a), and an estimated population growth of ca. 13 mg C m- 3 d- 1 for CI-CIII Calanus spp. and some older Pseudocalanus within the chl a maxima. Frontal waters were characterised by low chl a concentrations, but high abundances and production (around 1 g C m- 3 d- 1) of small copepods (Oithona spp.) and appendicularians (Fritillaria sp.). The estimated production of small-size zooplankton was an order of magnitude higher than the production of all other size groups combined, including large copepods. The high loss rates (- 166 to - 271 mg C m- 3 d- 1

System Studies for the ADTF: Target and Materials Test Station

International Nuclear Information System (INIS)

Cappiello, M.; Pitcher, E.; Pasamehmetoglu, K.

2002-01-01

To meet the objectives of the Advanced Accelerator Applications (AAA) program, the Accelerator-Driven Test Facility (ADTF) provides a world-class accelerator-driven test facility to: - Provide the capability to assess technology options for the transmutation of spent nuclear fuel and waste through a proof-of-performance. - Provide a user facility that allows testing of advanced nuclear technologies and applications, material science and research, experimental physics, and conventional nuclear engineering science applications. - Provide the capability, through upgrades or additions to the ADTF accelerator, to produce tritium for defense purposes, if required. - Provide the capability, through upgrades or additions, to produce radioisotopes for medical and commercial purposes. These missions are diverse and demand a facility with significant flexibility. In order to meet them, it is envisioned that we construct two target stations: the Target and Materials Test (TMT) station and the Subcritical Multiplier (SCM) test station. The two test stations share common hot-cell facilities for post-irradiation examination. It is expected the TMT will come online first, closely followed by the SCM. The TMT will provide the capability to: - Irradiate small samples of proposed ATW (accelerator-driven transmutation of waste) fuels and materials at prototypic flux, temperature, and coolant conditions (requires intense source of neutrons). - Perform transient testing. - Test liquid (lead-bismuth) and solid spallation targets with water, sodium, or helium coolant. - Test generation-IV fuels for advance nuclear systems (requires high-intensity thermal flux). - Irradiate fission product transmutation targets. - Test advanced fuel and coolant combinations, including helium, water, sodium, and lead-bismuth. - Produce isotopes for commercial and medical applications. - Perform neutron physics experiments. The SCM will provide the capability to: - Irradiate large samples of proposed ATW

Death qualification and prejudice: the effect of implicit racism, sexism, and homophobia on capital defendants' right to due process.

Science.gov (United States)

Butler, Brooke

2007-01-01

Two hundred venirepersons from the 12th Judicial Circuit in Bradenton, Florida completed the following measures: (1) one question that measured their level of support for the death penalty; (2) one question that categorized their death-qualification status; (3) 23 questions that measured their attitudes toward the death penalty (ATDP); (4) 22 questions that assessed their attitudes toward women (ATW); (5) 25 questions that measured their level of homophobia (H); (6) seven questions that assessed their level of modern racism (MR); (7) eight questions that measured their level of modern sexism (MS); and (8) standard demographic questions. Results indicated that as death-penalty support increased participants exhibited more positive attitudes toward the death penalty, more negative attitudes toward women, and higher levels of homophobia, modern racism, and modern sexism. Findings also suggested that death-qualified venirepersons exhibited more positive attitudes toward the death penalty and higher levels of homophobia, modern racism, and modern sexism. Finally, more positive attitudes toward the death penalty were correlated with more negative attitudes toward women and higher levels of homophobia, modern racism, and modern sexism. Legal implications are discussed. Copyright (c) 2007 John Wiley & Sons, Ltd.

Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods Integration and Evaluation Program (RMIEP)

International Nuclear Information System (INIS)

Payne, A.C. Jr.; Eide, S.A.; LaChance, J.C.; Whitehead, D.W.

1992-10-01

This volume presents the results of the initiating event and accident sequence delineation analyses of the LaSalle Unit II nuclear power plant performed as part of the Level III PRA being performed by Sandia National Laboratories for the Nuclear Regulatory Commission. The initiating event identification included a thorough review of extant data and a detailed plant specific search for special initiators. For the LaSalle analysis, the following initiating events were defined: eight general transients, ten special initiators, four LOCAs inside containment, one LOCA outside containment, and two interfacing LOCAs. Three accident sequence event trees were constructed: LOCA, transient, and ATWS. These trees were general in nature so that a tree represented all initiators of a particular type (i.e., the LOCA tree was constructed for evaluating small, medium, and large LOCAs simultaneously). The effects of the specific initiators on the systems and the different success criteria were handled by including the initiating events directly in the system fault trees. The accident sequence event trees were extended to include the evaluation of containment vulnerable sequences. These internal event accident sequence event trees were also used for the evaluation of the seismic, fire, and flood analyses

INCOGEN pre-feasibility study. Nuclear cogeneration

Energy Technology Data Exchange (ETDEWEB)

Van Heek, A.I.; De Haas, J.B.M.; Hogenbirk, A.; Klippel, H.T.; Kuijper, J.C.; Schram, R. [Netherlands Energy Research Foundation ECN, Petten (Netherlands); Hoogenboom, J.E.; Valko, J. [Interfaculty Reactor Institute IRI, Delft (Netherlands); Kanij, J.B.W.; Eendebak, B.T.; De Groot, P.C.; De Kler, R.C.F.; Stempniewicz, M.M. [KEMA, Arnhem (Netherlands); Van Dijk, A.B.; Bredman, B.; Van Essen, D.; Holtz, E.; Op `t Veld, R.; Tjemmes, J.G. [Stork Nucon, Amsterdam (Netherlands); Crommelin, G.A.K.; Crommelin-de Jonge, M.T. [eds.] [ROMAWA, Voorschoten (Netherlands)

1997-09-01

The Netherlands Programme to Intensify Nuclear Competence (PINK, abbreviated in Dutch) supported the technical and economical evaluation of a direct cycle High Temperature Reactor (HTR) installation for combined heat and power generation. This helium cooled, graphite moderated HTR based on the German HTR-M, is named INCOGEN (Inherently safe Nuclear COGENeration). The INCOGEN reference is a 40 MW HTR design by the US company Longmark Power International (LPI). The energy conversion system comprises a single-shaft helium turbine-compressor (2.3-1.0 MPa) directly coupled with a 16.5 MW generator, a recuperator and low-temperature (150C to 40C) heat exchangers (23 MW). Spherical fuel elements (60 mm diameter) will be added little by little, which keeps the core only marginally critical. Void core volume can accommodate added fuel for several years until defuelling. Analyses of failure scenarios (loss of coolant accident or LOCA, loss of flow accident or LOFA, anticipated transient without scram or ATWS) show no excess of maximum acceptable fuel temperature of 1600C. Scoping analyses indicate no severe graphite fires. Transient analyses of the turbine-compressor system indicate adequate control flexibility. Optimization and endurance testing of the helium turbine-compressor is recommended.

RF system considerations for large high-duty-factor linacs

International Nuclear Information System (INIS)

Lynch, M.T.; Ziomek, C.D.; Tallerico, P.J.; Regan, A.H.; Eaton, L.; Lawrence, G.

1994-01-01

RF systems are often a major cost item for linacs, but this is especially true for large high-duty-factor linacs (up to and including CW) such as the Accelerator for Production of Tritium (APT) or the Accelerator for Transmutation of nuclear Waste (ATW). In addition, the high energy and high average beam current of these machines (approximately 1 GeV, 100--200 mA) leads to a need for excellent control of the accelerating fields in order to minimize the possibility of beam loss in the accelerator and the resulting activation. This paper will address the key considerations and limitations in the design of the RF system. These considerations impact the design of both the high power RF components and the RF controls. As might be expected, the two concerns sometimes lead to conflicting design requirements. For example minimum RF operating costs lead to a desire for operation near saturation of the high power RF generators in order to maximize the operating efficiency. Optimal control of the RF fields leads to a desire for maximum overdrive capability in those same generators in order to respond quickly to disturbances of the accelerator fields

Design basis II: Design for events

International Nuclear Information System (INIS)

Frisch, W.

1982-01-01

In a lecture of this title, it could be expected that all events which are a basis for system and component design are described. According to the title of the Course 'Instrumentation and Control of Nuclear Power Plants' emphasis is put on events originating within the plant (no consideration of external events such as air plane crash or earth-quake). The lecture is divided into the two parts 'Transients' and 'Loss of coolant accidents (LOCAs)'. Due to the complex interaction between systems and components during transients, the first part is the main part of the lecture, while the second part (LOCAs) is only a very brief description of emergency core cooling system functions and the typical course of a large and small LOCA event. The first part on anticipated transients with intact primary coolant system boundary (non-LOCA-transients) covers several aspects of the analysis, such as classification, brief system description, transient description, analysis of anticipated transients without scram (ATWS) and analytical methods. Due to the time restriction necessary within the course, only a small section of the entire area can be presented in this paper. (orig.)

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

Development of integrated computer code for analysis of risk reduction strategy

International Nuclear Information System (INIS)

Kim, Dong Ha; Kim, See Darl; Kim, Hee Dong

2002-05-01

The development of the MIDAS/TH integrated severe accident code was performed in three main areas: 1) addition of new models derived from the national experimental programs and models for APR-1400 Korea next generation reactor, 2) improvement of the existing models using the recently available results, and 3) code restructuring for user friendliness. The unique MIDAS/TH models include: 1) a kinetics module for core power calculation during ATWS, 2) a gap cooling module between the molten corium pool and the reactor vessel wall, 3) a penetration tube failure module, 4) a PAR analysis module, and 5) a look-up table for the pressure and dynamic load during steam explosion. The improved models include: 1) a debris dispersal module considering the cavity geometry during DCH, 2) hydrogen burn and deflagration-to-detonation transition criteria, 3) a peak pressure estimation module for hydrogen detonation, and 4) the heat transfer module between the molten corium pool and the overlying water. The sparger and the ex-vessel heat transfer module were assessed. To enhance user friendliness, code restructuring was performed. In addition, a sample of severe accident analysis results was organized under the preliminary database structure

SBWR design update: Passively safe, nuclear power generation for the twenty first century

International Nuclear Information System (INIS)

Upton, H.A.; Torbeck, J.E.; Billig, P.F.; Duncan, J.D.; Herzog, M.

1996-01-01

This paper describes the current state of design, development and testing of a new generation of Boiling Water Reactors, the SBWR. The SBWR is a plant that will be significantly simpler to build, operate and maintain compared to operating plants. In this paper, the design and performance of the reference 670 MWe SBWR is summarized, the economics of SBWR power generation is addressed and the current developments in component testing and integrated system testing are given. This paper specifically discusses the current innovations and key reference design features of the SBWR including the RPV, depressurization system, pressure suppression system, flammability control system (based on passive autocatalytic recombiners), gravity driven cooling system, the passive containment cooling system, isolation condenser system and other unique engineered safety features that rely on gravity or stored energy to ensure core cooling, decay heat removal, and ATWS mitigation. The component and integrated system development testing summarized includes key results of recently concluded PANTHERS condenser tests conducted at SIET in Italy, GIRAFFE non-condensable gas testing by Toshiba in Japan, and the ongoing testing at the PANDA facility at PSI in Switzerland

Pore development in anodic alumina in sulphuric acid and borax electrolytes

International Nuclear Information System (INIS)

Garcia-Vergara, S.J.; Skeldon, P.; Thompson, G.E.; Habakaki, H.

2007-01-01

The formation of porous anodic films on an Al-3.5 at.%W alloy is compared in sulphuric acid and borax electrolytes in order to investigate pore development processes. The findings disclose that for anodizing in sulphuric acid, the pores develop mainly due to the influences of field-induced plasticity of the film and growth stresses; in borax, field-assisted dissolution dominates. The films formed in sulphuric acid are consequently much thicker than the layer of oxidized alloy and tungsten species are retained in the film. In contrast, with borax, the films and oxidized alloy layers are of similar thickness and tungsten species are lost to the electrolyte. Efficiencies of film growth are also significantly different, about 65% in sulphuric acid and about 52% in borax. The retention of tungsten species during anodizing in sulphuric acid is due to the localization of tungsten in the inner regions of the barrier layer and cell walls, with a layer of anodic alumina separating the tungsten-containing regions from the electrolyte. For borax, the tungsten is distributed more uniformly through the film material, enabling loss of tungsten species to the electrolyte from the pore base

ECN contributions to ADTT `96

Energy Technology Data Exchange (ETDEWEB)

Koning, A.J.

1996-07-01

An outline is presented of the status of nuclear data evaluation for accelerator-driven systems. The international effort consists mainly of measuring, compiling and calculating nuclear data for elements and isotopes relevant for transmutation of radioactive waste (ATW), energy amplification and other accelerator-related nuclear applications. We argue that input for global, macroscopic calculation schemes for hybrid nuclear systems basically should consist of three types of nuclear data: (a) High-energy transport codes for energies above about 150 MeV, (b) neutron and proton transport data files for energies below about 150 MeV and (c) neutron and proton transmutation/activation libraries below about 150 MeV. Our specific contribution to the field concerns (b) and (c). The progress of the evaluation of high-energy nuclear data files for the most important materials and the related compilation of nuclear reaction information is reported. The evaluated data are calculated with the computer codes ECIS95, MINGUS and GNASH and are stored in ENDF6-format. We illustrate the library production with a short outline of the employed physical methods. Finally, we briefly discuss the application of the activation/transmutation library ECNAF96. (orig.).

Technologies for destruction of long-lived radionuclides in high-level nuclear waste: Overview and requirements

International Nuclear Information System (INIS)

Arthur, E.D.

1993-01-01

This paper, and this topical session on Nuclear Waste Minimization, Management and Remediation, focuses on two nuclear systems, and their associated technologies, that have the potential to address concerns surrounding long-lived radionuclides in high-level waste. Both systems offer technology applicable to HLW from present light-water reactors (LWR). Additionally these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes, long-term fuel supplies, and enhanced safety. The first system is the Integral Fast Reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while simultaneously burning long-lived actinide waste. IFR applications include burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, Accelerator Transmutation of Waste (ATW), is based upon an accelerator-induced intense source of thermal neutrons and is aimed at destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

An innovative design approach to a cost effective commercial liquid metal reactor

International Nuclear Information System (INIS)

Carelli, M. D.; Garkisch, H. D.; Hundal, R.; Arie, K.; Handa, N.; Ota, H.; Matsuyama, S.; Todreas, N.; Hejzlar, P.; Wells, P.; Louis, T. S.

2008-01-01

use the SGHARS (Steam Generator Auxiliary Removal System) which uses the steam generator. In the abnormal event of the loss of feedwater we can use the RVACS (Reactor Vessel Auxiliary Cooling System). In a pool configuration, decay heat is transferred by conduction to the vessel and then by radiation and conduction to the guard vessel. The guard vessel is cooled initially by evaporative cooling followed by natural draft air cooling of the external surface. This approach is similar to the one taken for the AP1000. Finally, another innovative feature is the inclusion of a tertiary shutdown system. Thanks to the inherent reactivity feedbacks response the LMR can be brought to hot shutdown even when both the primary and secondary shutdown systems are unavailable. However, the core will experience high transient temperatures, before they are finally brought down by the reactivity feedbacks. The designer has to account for these high temperatures attained during ATWS (Anticipated Transients Without Scram). Since an ATWS is by definition the absence of both the primary and secondary shutdown systems, adoption of a tertiary system will eliminate the ATWS from design considerations. The system considered is different from the other two shutdown systems both in terms of absorber configuration and mode of insertion. The probability of simultaneous failure of all three diverse systems is less than the 10-8 threshold. Elimination of ATWS design considerations allows a power increase of the order of 10%, while the capital cost increase associated with the tertiary system is less than 1%. In summary our approach to achieve economic competitiveness with advanced LWRs is very simple: design the ARR with the same philosophy and within similar constraints. This is implemented by the innovative components presented here. (authors)

Zirconium and technetium recovery and partitioning in the presence of actinides in modified Purex process for ATW program. Final report

International Nuclear Information System (INIS)

Dzekun, E.G.; Fedorov, Y.S.; Galkin, B.Y.; Lyubtsev, R.I.; Mashkin, A.N.; Mishin, E.N.; Zilberman, B.Y.

1994-01-01

The modified Purex process flowsheet is based on combination of all irradiated materials, their joint dissolution and reprocessing as a NPP spent fuel solution with abnormal Pu content after addition of recycled depleted U concentrate. Some groups of long-lived radionuclides could be completely recovered and localized at the stage of extraction reprocessing using 30% TBP. Studies were conducted for 10 y to develop the process for recovery, concentration, and localization of U, Pu, Np, Tc, and Zr within 1st extraction cycle. Actinides are recovered from high-level raffinate of this cycle after evaporation and feed adjustment. Results in this report show that combined deep recovery of several elements from highly irradiated materials by TBP extraction, for further transmutation, is possible. Selective stripping of Zr from solvent phase containing U, Pu, Np, and Tc is quite effective. Development of the modified Purex process is not complete; main problem to be solved should be oxide separation from the loop and permissible storage duration before reprocessing and reuse in the loop

Zirconium and technetium recovery and partitioning in the presence of actinides in modified Purex process for ATW program. Final report

Energy Technology Data Exchange (ETDEWEB)

Dzekun, E.G.; Fedorov, Y.S.; Galkin, B.Y.; Lyubtsev, R.I.; Mashkin, A.N.; Mishin, E.N.; Zilberman, B.Y. [Radievyj Inst., Leningrad (Russian Federation)

1994-12-31

The modified Purex process flowsheet is based on combination of all irradiated materials, their joint dissolution and reprocessing as a NPP spent fuel solution with abnormal Pu content after addition of recycled depleted U concentrate. Some groups of long-lived radionuclides could be completely recovered and localized at the stage of extraction reprocessing using 30% TBP. Studies were conducted for 10 y to develop the process for recovery, concentration, and localization of U, Pu, Np, Tc, and Zr within 1st extraction cycle. Actinides are recovered from high-level raffinate of this cycle after evaporation and feed adjustment. Results in this report show that combined deep recovery of several elements from highly irradiated materials by TBP extraction, for further transmutation, is possible. Selective stripping of Zr from solvent phase containing U, Pu, Np, and Tc is quite effective. Development of the modified Purex process is not complete; main problem to be solved should be oxide separation from the loop and permissible storage duration before reprocessing and reuse in the loop.

A study on reactor core failure thresholds to safety operation of LMFBR

International Nuclear Information System (INIS)

Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

2006-01-01

Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

A study on the regulatory approach of major technical issues

Energy Technology Data Exchange (ETDEWEB)

Chang, Keun Sun; Oh, S. H.; Kang, H. J.; Kim, G. S. [Sunmoon Univ., Asan (Korea, Republic of); Lee, S. H.; Baek, W. P.; Yang, S. H.; Jeong, Y. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

1999-02-15

This project is to provide the regulatory direction of 4 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are are as follows : related to the classification and acceptance criteria of plant conditions, currently used classification and safety requirement were reviewed and regulatory direction was proposed. Among multiple failures, it is identified that SBO, TLOFW, multiple SGTR and ATWS are basically to be considered for additional requirements for advanced reactors. This study reviewed risk aspects, design consideration, and trends of safety requirements, and proposed fundamental safety requirements to be applied for KNGR. Multiple steam generator tube failure is a significant safety concern because of the possibility of release of radionuclides to the environment through containment bypass. Proposed safety requirement for this event can be categorized mainly as analysis requirement, design evaluation requirement and PSA requirement; For protection of containment failure, a reasonable safety position is necessary through and integrated review of possibility of severe accident occurrence, effects of sever accident mitigation features and cost effects of these design features. With this consideration safety requirements developed are the analysis requirement, provision of protective measures and survivability/availability of protective measures.

ALMOD-JRC computer program

International Nuclear Information System (INIS)

Cacciabue, P.C.; Lisanti, B.; Tozzi, A.

1984-01-01

This paper discusses the details concerning the newly developed or modified models of the computer program ALMOD-JRC, originating from ALMOD 3/Rel 4. The most important argument for the implementation of the new models was the need to enlarge the spectrum of the simulated phenomena, and to improve the simulation of experimental facilities such as LOFT or LOBI. This has led to a better formulation of the heat transfer and pressure drops correlations and to the implementation of the treatment of the heat losses to structural materials. In particular a series of test cases on real power plants, a pre-test examination of a LOBI station blackout ATWS experiment and the post test analysis of the L9-3 experiment, show the ability of ALMOD-JRC to correctly simulate PWR incident sequences. Although in ALMOD-JRC the code capabilities have been expanded, the limitations of the original version of the program still hold for what concerns the treatment of the coolant thermohydraulics as homogeneous flow for the two phase conditions in the primary coolant circuit. The other interesting feature of the new code is the remarkably shorter running times obtained with the introduction of simplified numerical treatments for the solving equations, without significant loss of accuracy of results

Transport ac loss studies of YBCO coated conductors with nickel alloy substrates

International Nuclear Information System (INIS)

Duckworth, R C; Thompson, J R; Gouge, M J; Lue, J W; Ijaduola, A O; Yu, D; Verebelyi, D T

2003-01-01

Transport alternating current (ac) loss measurements were performed on a series of rolling-assisted biaxially textured substrate (RABiTS) processed YBa 2 Cu 3 O x (YBCO) coated conductors at 77 K. While each sample possessed a 1 μm layer of YBCO and a 3 μm silver cap layer, two different nickel alloy substrates were used and their impact on the ac loss was examined. Both substrates possessed a 75 μm Ni-5 at%W base, but one substrate also had a 2 μm nickel overlayer as part of the buffer layer architecture. The ac losses, which were determined by thermal and electrical measurements, contained two dominant contributions: superconductive hysteresis in the YBCO and ferromagnetic hysteresis in the substrates. The superconductive component followed the Norris elliptic model for the substrate with the nickel overlayer and the Norris thin strip model for the substrate without the nickel overlayer. The substrates' ferromagnetic loss was determined separately through magnetization measurements, which showed that this loss contribution was independent of the presence of the nickel overlayer for effective ac currents less than 50 A. While the overall loss was lower for the thin-strip-like conductor with no nickel overlayer, further research is necessary to strengthen this connection

Nuclear research and nuclear technology in the Federal Republic of Germany

International Nuclear Information System (INIS)

Anon.

1981-01-01

The atomwirtschaft-atomtechnik has reflected the development of this quarter century. In this jubilee edition it describes the future lines of development. It has invited the Federal German companies and institutions of the branch to present their performance potential in the form of monography - more detailed than usually. This invitation was accepted by 81 of the most important enterprises. The figure also includes a number of important service companies, the research centres of the country, and last not least, a number of energy supply enterprises. Part 2 of this jubilee edition as a whole offers a crossection of the present performances offered in the German nuclear research, nuclear techniques, and the planning and service belonging to nuclear power operation. For the English-speaking readers, a digest part was set up in part 3 of the present edition. In part 4, the reader will find a product index in German and English. Each key-word indicates an offering firm by the page number allocated. Access to the monographies (part 2) and the digest (part 3) can be found in the listing of the monography-advertisers from page 102 on. The atw-jubilee edition closes with part 5, with product advertisements of companies from home and abroad. (orig./UA) [de

Review of EU-APR Design for Selected Safety Issues of WERNA RHWG 2013

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Kim, Ji Hwan [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Western European Nuclear Regulators' Association (WENRA) was established in 1999 to develop a harmonized approach to nuclear safety and radiation protection and their regulation. In 2013, the Reactor Harmonization Working Group (RHWG) of WENRA sets out the common positions on the seven selected key safety issues. This paper is to introduce the regulatory positions of WENRA RHWG 2013 and to review the compliance of the EU-APR with them. In this paper, we reviewed the compliance of the EUAPR regarding seven safety issues for new NPPs presented by WERNA RHWG in 2013. The EU-APR design fully complies with all WERNA RHWG safety issues since the following measures have been incorporated in it: - Successive five levels of DiD maintaining independence between different levels of DiD - Diverse design against multiple failure events such as ATWS, SBO, Loss of Ultimate Heat Sink, and Loss of Spent Fuel Pool Cooling - SAs dedicated mitigation systems to ensure the containment integrity during the SAs. - Practically eliminates accident sequences with a large or early release of radiological materials by diverse designs for multiple failure events, SAs dedicated mitigation system, and double containment design - Standard site parameters not lead to core melt accidents due to natural or man-made external hazards.

Czynniki prognostyczne progresji radiologicznej w reumatoidalnym zapaleniu stawów

Directory of Open Access Journals (Sweden)

Piotr Wiland

2010-08-01

Full Text Available Możliwość przewidywania odległych konsekwencji reumatoidalnegozapalenia stawów ma zasadnicze znaczenie dla podejmowaniawłaściwych decyzji terapeutycznych u danego chorego. Markeryprognostyczne są pomocne w identyfikacji chorych o dużym ryzykuszybkiej progresji radiologicznej. Utwierdzają one lekarzaw decyzji o rozpoczęciu intensywnego leczenia u chorych z możliwąznaczną destrukcją stawów w ciągu następnych kilku lat.W artykule zaprezentowano pilotażowy model ryzyka dla przewidywaniaszybkiej progresji radiologicznej. W celu stworzenia tegomodelu posłużono się danymi, w tym wynikami badań radiologicznychpochodzącymi z badania ASPIRE, w którym porównywanomonoterapię metotreksatem z leczeniem skojarzonym metotreksatemi infliksymabem. W tym modelu wybrano wyjściowe parametry,które są łatwe do uzyskania w rutynowej praktyce klinicznej,takie jak: stężenie białka C-reaktywnego, wartość odczynuopadania krwinek czerwonych, liczbę obrzękniętych stawów orazmiano czynnika reumatoidalnego. Ten macierzowy model ryzykamoże być przydatny w ocenie ryzyka postępującego uszkodzeniastawów, szczególnie u chorych z wczesnym zapaleniem stawów.

An investigation of the applicability of the new ion exchange resin, Reillex{trademark}-HPQ, in ATW separations. Milestone 4, Final report

Energy Technology Data Exchange (ETDEWEB)

Ashley, K.R.; Ball, J.; Grissom, M.; Williamson, M.; Cobb, S.; Young, D.; Wu, Yen-Yuan J.

1993-09-07

The investigations with the anion exchange resin Reillex{trademark}-HPQ is continuing along several different paths. The topics of current investigations that are reported here are: The sorption behavior of chromium(VI) on Reillex{trademark}-HPQ from nitric acid solutions and from sodium hydroxide/sodium nitrate solutions; sorption behavior of F{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; sorption behavior of Cl{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; sorption behavior of Br{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; and the Honors thesis by one of the students is attached as Appendix II (on ion exchange properties of a new macroperous resin using bromide as the model ion in aqueous nitrate solutions).

Behaviour of a PWR with core protection system (SSN) in case of accidents due to power failure, ATWS and steam generator rupture

International Nuclear Information System (INIS)

Boncompagni, S.; Fulceri, P.; Oriolo, F.

1985-01-01

The results of the analysis of the transient fallowing internal and external power failure, without scram, in the nuclear power plant of the Italian Unified Nuclear Project are examined. The availability of ECCS is excluded while the breakage of a tube in each steam generator is supposed, togheter with the presence of an original safety system known as SSN (core protection system). Computations have been performed by using Mark 6 RELAP4 code. The study of the transient and the physical model used are briefly illustrated. Finally the results achieved are analysed

Development of Integrated Evaluation System for Severe Accident Management

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

2007-06-15

The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

Mixing of high density solution in vertical upward flow

International Nuclear Information System (INIS)

Kumamaru, Hiroshige; Hosogi, Nobuyoshi; Komada, Toshiaki; Fujiwara, Yoshiki

1999-01-01

Experimental and analytical studies have been performed in order to provide fundamental data and a numerical calculation model on the mixing of boric acid solution, injected from the standby liquid control system (SLCS), under a low natural circulation flow during an ATWS in a BWR. First, fundamental experiments on the mixing of high-density solution in vertically-upward water flow have been performed by using a small apparatus. Mixing patterns observed in the experiments have been classified to two groups, i.e. complete mixing (entrainment) and incomplete mixing (entrainment). In the complete mixing, the injected high-density solution is mixed (entrained) completely into the vertically-upward water flow. From the experiments, the minimum water flow rates in which the complete mixing (entrainment) is achieved have been obtained for various solution densities and solution injection rates. Secondly, two-dimensional numerical calculations have been performed. A continuity equation for total fluid, momentum equations in two directions and a continuity equation for solute are solved by using the finite difference method for discretization method and by following the MAC method for solution procedure. The calculations have predicted nearly the minimum water flow rate in which the complete mixing is achieved, while the calculations have been performed only for one combination of the solution density and solution injection rate until now. (author)

The 50th anniversary of the N.S. Otto Hahn. When nuclear power said 'Ahoy'

International Nuclear Information System (INIS)

Reinartz, Jerome

2013-01-01

In September 1963 the construction of the only German nuclear ship to date, NS Otto Hahn, started in Kiel. For the 50 th anniversary, atw remembers an important part of nuclear technology history in Germany: The research and freight ship shows one thing above all in retrospect: The technology ran reliably. But cost pressure and reservations shattered the dream of a nuclear power shipping era. Until it was decommissioned in 1979 the ship travelled a total of 650,000 sea miles and called at 33 harbours in 22 countries. On the research level, the 'Otto Hahn' could satisfy expectations, however, it could not ring in an era of nuclear shipping - the atomic boat could never cover its operating costs with its freight trips and permission to call at foreign ports were rare. However, on the one hand, the ship's journeys, sometimes under hard weather conditions, demonstrated just how robust and durable the 'progressive pressurized water reactor' on board was, on the other hand, the 'Otto Hahn' had by all means been a prototype which under other market conditions could have been a model for nuclear container ships. In any case, it proved the performance capacity of the then still young German and European nuclear technology industry, that did not need to hide behind the Russian and American competition. (orig.)

Is a 4-train support ''always'' more reliable than a 2-train support

International Nuclear Information System (INIS)

Guey, C.N.; Arrieta, L.; Youngblood, R.

1986-06-01

Once the gross features of a frontline fluid system have been defined, one must consider what support system configuration will provide the best overall system performance. This paper considers different dc bus configurations for a given emergency feedwater system. Results indicate that a four-train support system (i.e., 4 dc buses) gives a lower system unavailability for transients, but a higher system unavailability for Anticipated Transients Without Scram (ATWS), than a two-train support system (i.e., two dc buses). This serves to illustrate that more trains do not necessarily provide higher reliability, and that a configuration choice which is better for one mission success criterion may be worse for another. Because of the small characteristic unreliability of dc buses, the numerical comparisons made here are not dramatic, but the underlying topological point is nevertheless broadly applicable. The EFWS selected in this study consists of two turbine-driven pumps, and two motor-driven pumps, with associated piping, instruments, and valves. There are two divisions of EFWS, each division including one turbine-driven pump train, and one motor-driven pump train with a crosstie that can provide an alternate flow path in case one train is not available. Each of the motor-driven pumps has one ac power bus associated with it. No ac power is required for the turbine-driven trains

Using proliferation risk as a design metric in the development of nuclear systems

International Nuclear Information System (INIS)

Beard, C.; Lebouf, R.

2001-01-01

The necessity has arisen for newly proposed nuclear systems to be evaluated with regard to their potential aid to any proliferation. Thus, a mechanism is needed to introduce nonproliferation as a measure in the design phase of a new nuclear system. To accomplish this, a methodology for quantifying and measuring the proliferation risk of proposed system options is required. Such quantification has its difficulties due to inherent uncertainty, e.g. what is the probability that a quantity of material will be stolen in a given situation? Also, the lack of data on such occurrences makes the task of quantification nearly insurmountable. A systematic approach is necessary to estimate the proliferation risk. Currently, an advanced nuclear power system, the Accelerator Transmutation of Waste (ATW) program has been initiated to develop a system that will concurrently generate electricity while destroying long-lived radioactive isotopes. Therefore, because of the issues noted above, an effort to introduce proliferation risk into the design phase has been started. The purpose of this paper is to review previous work in quantification of proliferation risk in an effort to develop the proper basis for the current work. It should be noted that while proliferation on a national level has been studied extensively, efforts to quantify proliferation risk of individual nuclear systems or processes have been limited. Consequently, the available literature base is relatively sparse. (author)

Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

International Nuclear Information System (INIS)

Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L.

1989-08-01

This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs

Using proliferation risk as a design metric in the development of nuclear systems

Energy Technology Data Exchange (ETDEWEB)

Beard, C.; Lebouf, R. [Texas Univ., Austin, TX (United States). Nuclear Engineering Teaching Lab.

2001-07-01

The necessity has arisen for newly proposed nuclear systems to be evaluated with regard to their potential aid to any proliferation. Thus, a mechanism is needed to introduce nonproliferation as a measure in the design phase of a new nuclear system. To accomplish this, a methodology for quantifying and measuring the proliferation risk of proposed system options is required. Such quantification has its difficulties due to inherent uncertainty, e.g. what is the probability that a quantity of material will be stolen in a given situation? Also, the lack of data on such occurrences makes the task of quantification nearly insurmountable. A systematic approach is necessary to estimate the proliferation risk. Currently, an advanced nuclear power system, the Accelerator Transmutation of Waste (ATW) program has been initiated to develop a system that will concurrently generate electricity while destroying long-lived radioactive isotopes. Therefore, because of the issues noted above, an effort to introduce proliferation risk into the design phase has been started. The purpose of this paper is to review previous work in quantification of proliferation risk in an effort to develop the proper basis for the current work. It should be noted that while proliferation on a national level has been studied extensively, efforts to quantify proliferation risk of individual nuclear systems or processes have been limited. Consequently, the available literature base is relatively sparse. (author)

Development of Integrated Evaluation System for Severe Accident Management

International Nuclear Information System (INIS)

Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y.

2007-06-01

The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs

Interwencje redukujące zagrożenia stereotypem w środowisku edukacyjnym

Directory of Open Access Journals (Sweden)

Dariusz Drążkowski

2014-09-01

Full Text Available Zagrożenie stereotypem wpływa na osiąganie gorszych wyników w nauce przez uczniów należących do grup negatywnie stereotypizowanych. U uczniów z mniejszości etnicznych lub rasowych wpływa ono destrukcyjnie nie tylko na osiągnięcia szkolne, ale także na poziom szczęścia czy stan zdrowia. Istnieje więc uzasadnione zapotrzebowanie na metody walki z tym zjawiskiem, które w ostatnich latach intensywnie starają się zaspokoić badania prowadzone przez psychologów społecznych. Celem artykułu jest przedstawienie i systematyzacja wyników badań na temat interwencji redukujących zagrożenie stereotypem w środowisku edukacyjnym. Badania te uporządkowano wokół interwencji ukierunkowanych na społeczeństwo i szkołę, interwencji związanych z zachowaniem nauczycieli i z warunkami przeprowadzania egzaminów testowych oraz skoncentrowanych na uczniu. Wyniki omówionych badań nasuwają efektywne i często łatwe do zastosowania metody zmniejszające zagrożenie stereotypem wśród uczniów. Interwencje oparte na tych metodach mogą wspierać tworzenie wyrównanych warunków nauki dla każdego ucznia.

The 50{sup th} anniversary of the N.S. Otto Hahn. When nuclear power said 'Ahoy'; Als die Kernenergie 'Ahoi' rief. NS Otto Hahn. Nuklearbetriebenes Forschungs- und Frachtschiff

Energy Technology Data Exchange (ETDEWEB)

Reinartz, Jerome [atw-Redaktion, Hattingen (Germany)

2013-11-15

In September 1963 the construction of the only German nuclear ship to date, NS Otto Hahn, started in Kiel. For the 50{sup th} anniversary, atw remembers an important part of nuclear technology history in Germany: The research and freight ship shows one thing above all in retrospect: The technology ran reliably. But cost pressure and reservations shattered the dream of a nuclear power shipping era. Until it was decommissioned in 1979 the ship travelled a total of 650,000 sea miles and called at 33 harbours in 22 countries. On the research level, the 'Otto Hahn' could satisfy expectations, however, it could not ring in an era of nuclear shipping - the atomic boat could never cover its operating costs with its freight trips and permission to call at foreign ports were rare. However, on the one hand, the ship's journeys, sometimes under hard weather conditions, demonstrated just how robust and durable the 'progressive pressurized water reactor' on board was, on the other hand, the 'Otto Hahn' had by all means been a prototype which under other market conditions could have been a model for nuclear container ships. In any case, it proved the performance capacity of the then still young German and European nuclear technology industry, that did not need to hide behind the Russian and American competition. (orig.)

MOD approach for the growth of epitaxial CeO2 buffer layers on biaxially textured Ni-W substrates for YBCO coated conductors

International Nuclear Information System (INIS)

Bhuiyan, M S; Paranthaman, M; Sathyamurthy, S; Aytug, T; Kang, S; Lee, D F; Goyal, A; Payzant, E A; Salama, K

2003-01-01

We have grown epitaxial CeO 2 buffer layers on biaxially textured Ni-W substrates for YBCO coated conductors using a newly developed metal organic decomposition (MOD) approach. Precursor solution of 0.25 M concentration was spin coated on short samples of Ni-3 at%W (Ni-W) substrates and heat-treated at 1100 C in a gas mixture of Ar-4%H 2 for 15 min. Detailed x-ray studies indicate that CeO 2 films have good out-of-plane and in-plane textures with full-width-half-maximum values of 5.8 deg. and 7.5 deg., respectively. High temperature in situ XRD studies show that the nucleation of CeO 2 films starts at 600 C and the growth completes within 5 min when heated at 1100 C. SEM and AFM investigations of CeO 2 films reveal a fairly dense microstructure without cracks and porosity. Highly textured YSZ barrier layers and CeO 2 cap layers were deposited on MOD CeO 2 -buffered Ni-W substrates using rf-magnetron sputtering. Pulsed laser deposition (PLD) was used to grow YBCO films on these substrates. A critical current, J c , of about 1.5 MA cm -2 at 77 K and self-field was obtained on YBCO (PLD)/CeO 2 (sputtered)/YSZ (sputtered)/CeO 2 (spin-coated)/Ni-W

Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

2000-10-01

In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

Preliminary Evaluation of the Diverse Protection System in PGSFR

International Nuclear Information System (INIS)

Jeong, Taekyeong; Chang, Won-Pyo; Seong, Seung Hwan; Ahn, Sang June; Kang, Seok Hun; Choi, Chiwoong; Yoo, Jin; Lee, Kwi Lim; Lee, Seung Won; Jeong, Jae-Ho; Ha, Kwi-Seok

2015-01-01

The anticipated transient without scram (ATWS) is defined as an abnormal transient with failure of scram actuation. It is one of the “worst case” accident based on the United States Nuclear Regulatory Commission (U.S.NRC). Consideration frequently motivates the NRC to take regulatory action. An evaluation of this event is also a general requirement due to a potential safety issue that may lead to core damage under postulated condition. This paper estimated the set-points sensitivity test of the diverse protection system (DPS) related with unprotected events of the prototype generation-IV sodium cooled fast reactor (PGSFR) including unprotected transient over power (UTOP) and unprotected loss of flow (ULOF) by MARS-LMR code. The variation of the power to flow (P/Q) of UTOP and ULOF is illustrated to conduct the set-points sensitivity test of DPS. Also we estimated the effect of the DPS introduction after selecting UTOP, ULOF event as the unprotected events which are predicted to aggravate the events. This paper estimated the set-points sensitivity test of DPS related with unprotected events of PGSFR including UTOP and ULOF by MARS-LMR code. The results indicated that there is no significant difference in both RPS and DPS about the initiating time of each event. Therefore, this study found that the urgent manage for safety of the reactor when RPS failed is possible by the applying DPS

Dependability Evaluation of Advanced Diverse Protection System

International Nuclear Information System (INIS)

Oh, Yang Gyun; Lee, Yoon Hee; Sohn, Se Do; Baek, Seung Min; Lee, Sang Jeong

2014-01-01

For the mitigation of anticipated transients without scram (ATWS) as well as common cause failure (CCF) within the plant protection system (PPS) and the emergency safety feature . component control system (ESF-CCS), the diverse protection system (DPS) has been designed by KEPCO Engineering and Construction Company. Recently KEPCO E and C has developed the advanced diverse protection system (ADPS), which has four redundant channels, in an attempt to enhance a fault-tolerant capability of the system. For the evaluation of overall system improvement effects of the ADPS compared with the DPS, the dependability evaluation results are described herein. For all dependability attributes, this paper suggests a practical dependability evaluation method which uses quantitative dependability scores and indices. An overall dependability evaluation index (DEI) for the ADPS is evaluated with the average value of reliability/ security/maintainability/safety indices (i.e., RID, SID, MID, and SID') for dependability. The evaluation results show that the DEI value of ADPS can be improved by approximately 23% compared with that of the DPS, thanks to its fault-tolerant system architecture, software design changes, and external interface design features. Several suggestions have been made, in this paper, of an overall quantitative dependability evaluation method for the nuclear instrumentation and control (I and C) systems including the DPS and ADPS, and the usefulness of dependability evaluation on nuclear I and C systems has been confirmed

Nuclear power plants in Europe 1996. Report on operation, construction and planning in 18 European countries

International Nuclear Information System (INIS)

Anon.

1996-01-01

Eighteen European countries operate and build, respectively, nuclear power plants in 1996. The 'Nuclear Power Plants in Europe 1996' atw report singles out the main events and lines of development. As per August 1996, 216 (1995: 215) nuclear generating units (which means power reactors for the purposes of this report) with an aggregate 177,916 (177,010) MWe installed gross capacity are in operation and 24 (26) units with 23,086 (24,786) MWe are under construction in seven countries. This adds up to a total of 240 (241) nuclear generating units with an aggregate 201,002 (201,796) MWe. In the nuclear power plants in Europe, some 1067 (1048) TWh of nuclear power was converted into electric power in 1995; 802 (792) TWh of this aggregate was converted in 142 (137) units in the European Union (EU). In the EU the share of nuclear power in the public supply of electricity was 36%. Lithuania, with 86 (77)% has the highest share of nuclear power in Europe, followed by France with 76 (75)% and Belgium with 56%. The lowest percentage, only 4 (5)%, is recorded in the Netherlands. As a consequence of electricity imports, nuclear power holds considerable shares in the public electricity supply also of countries in which no nuclear power are operated, such as Italy or Austria. (orig.) [de

Nuclear power plants in Europe 1995. Report about operation, construction, and planning in 18 European countries

International Nuclear Information System (INIS)

Anon.

1995-01-01

Report about Operation, Construction, and Planning in 18 European Countries Eighteen European countries operate and build, respectively, nuclear power plants in 1995. The ''Nuclear Power Plants in Europe 1995'' atw report singles out the main events and lines of development. As per August 1995, 214 (1994: 215) nuclear generating units (which means power reactors for the purposes of this report) with an aggregate 177,010 (176,322) MWe installed gross capacity are in operation in seventeen countries, and 26 (30) units with 24,786 (28,086) MWe are under construction in seven countries. This adds up to a total of 240 (245) nuclear generating units with an aggregate 201,796 (204,408) MWe. In the nuclear power plants in Europe, some 1048 TWh of nuclear power was converted into electric power in 1994; 792 TWh of this aggregate was converted in 137 units in the European Union (EU). In the EU the share of nuclear power in the public supply of electricity was 36%. Lithuania, with 77%, has the highest share of nuclear power in Europe, followed by France with 75% and Belgium with 56%. The lowest percentage, only 5%, is recorded in the Netherlands. As a consequence of electricity imports, nuclear power holds considerable shares in the public electricity supply also of countries in which no nuclear power plants are operated, such as Italy or Austria. (orig.) [de

Preliminary Evaluation of the Diverse Protection System in PGSFR

Energy Technology Data Exchange (ETDEWEB)

Jeong, Taekyeong; Chang, Won-Pyo; Seong, Seung Hwan; Ahn, Sang June; Kang, Seok Hun; Choi, Chiwoong; Yoo, Jin; Lee, Kwi Lim; Lee, Seung Won; Jeong, Jae-Ho; Ha, Kwi-Seok [KAERI, Daejeon (Korea, Republic of)

2015-05-15

The anticipated transient without scram (ATWS) is defined as an abnormal transient with failure of scram actuation. It is one of the “worst case” accident based on the United States Nuclear Regulatory Commission (U.S.NRC). Consideration frequently motivates the NRC to take regulatory action. An evaluation of this event is also a general requirement due to a potential safety issue that may lead to core damage under postulated condition. This paper estimated the set-points sensitivity test of the diverse protection system (DPS) related with unprotected events of the prototype generation-IV sodium cooled fast reactor (PGSFR) including unprotected transient over power (UTOP) and unprotected loss of flow (ULOF) by MARS-LMR code. The variation of the power to flow (P/Q) of UTOP and ULOF is illustrated to conduct the set-points sensitivity test of DPS. Also we estimated the effect of the DPS introduction after selecting UTOP, ULOF event as the unprotected events which are predicted to aggravate the events. This paper estimated the set-points sensitivity test of DPS related with unprotected events of PGSFR including UTOP and ULOF by MARS-LMR code. The results indicated that there is no significant difference in both RPS and DPS about the initiating time of each event. Therefore, this study found that the urgent manage for safety of the reactor when RPS failed is possible by the applying DPS.

Chemical separations schemes for partitioning and transmutation systems

International Nuclear Information System (INIS)

Laidler, J.

2002-01-01

In the initial phase of the U.S. Accelerator Transmutation of Waste (ATW) program, a single-tier system was foreseen in which the transuranics and long-lived fission products (specifically, 99 Tc and 129 I) recovered from spent LWR oxide fuel would be sent directly to an accelerator-driven transmuter reactor [1]. Because the quantity of fuel to be processed annually was so large (almost 1,500 tons per year), an aqueous solvent extraction process was chosen for LWR fuel processing. Without the need to separate transuranics from one another for feed to the transmuter, it became appropriate to develop an advanced aqueous separations method that became known as UREX. The UREX process employs an added reagent (acetohydroxamic acid) that suppresses the extraction of plutonium and promotes the extraction of technetium together with uranium. Technetium can then be efficiently removed from the uranium; the recovered uranium, being highly decontaminated, can be disposed of as a low-level waste or stored in an unshielded facility for future use. Plutonium and the other transuranic elements, plus the remaining fission products, are directed to the liquid waste stream. This stream is calcined, converting the transuranics and fission products to their oxides. The resulting oxide powder, now representing only about four percent of the original mass of the spent fuel, is reduced to metallic form by means of a pyrometallurgical process. Subsequently, the transuranics are separated from the fission products in another pyro-metallurgical step involving molten salt electrorefining

Implementation of digital control and protection systems of China advanced research reactor

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

Generalized design formulas for low energy electromagnetic quads

International Nuclear Information System (INIS)

Liska, D.J.

1994-06-01

This technical note is the result of the quadrupole magnet design efforts that went into the development of proposals for large high-powered linear accelerators such as the Accelerator for Production of Tritium (APT), Accelerator for Base Conversion (of Plutonium) (ABC), and Accelerator for Treatment of (radioactive) Waste (ATW). In all these applications it was necessary to develop designs for numerous (hundreds) of electromagnetic quadrupoles (EMQs). EMQs are required since long-term reliability, radiation damage potential, and large aperture dictate against the use of permanent magnet quadrupoles (PMQs) for these powerful machines. One object of the magnet design effort was to provide a quick, reliable, and easy means of converting raw physics requirements (magnetic impulse, focal length, and boretube aperture) into realistic electrical, cooling, facility interface, and mechanical specifications and configurations--in other words, to easily convert physics requirements to a reliable design that could be drawn on paper, shown to vendors, and presented to peer review committees as a well-developed and believable concept. The empirical formulas that were derived have been gathered together in this technical note. They will be useful for other designers interested in an easy way of coming up with a rather complete mechanical as well as electrical and magnetic design for EMQs. Included are lab tests of designs derived from these formulas and comparisons with other real EMQ designs. These demonstrate the good accuracy of the empirical formulas

Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

International Nuclear Information System (INIS)

Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

2007-01-01

Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

YKAe - Research programme on nuclear power plant systems behaviour and operational aspects of safety

International Nuclear Information System (INIS)

Mattila, L.; Vanttola, T.

1992-01-01

The major part of nuclear energy research in Finland has been organised as five-year nationally coordinated research programs. The research programme on Systems Behaviour and Operational Aspects of Safety is under way during 1990-1994. Its annual volume has been about 35 person-years and its annual expenditure about FIM 18 million. Studies in the field on safe operational margins of nuclear fuel and reactor core concentrate on fuel high burn-up behaviour, VVER fuel experiments, and reactor core behaviour in complex reactivity transients such as 3-D phenomena and ATWS events. The PACTEL facility is used for the thermal hydraulic studies of the Loviisa type reactors (scaled 1:305). Validation of accident analysis codes is carried out by participation in international standard problems. Advanced foreign computer codes for severe reactor accidents are implemented, modified as needed and applied to level-2 PSAs and the improvement of accident management procedures. Fire simulation methods are tested using data from experiments in the German HDR facility. A nuclear plant analyzer for efficient safety analyses is being developed using the APROS process simulation environment. Computerized operator support systems are being studied in cooperation with the OECD Halden Project. The basic factors affecting plant operator activities and the development of their competence are being investigated. A comprehensive system for the control of plant operational safety is being developed by combining living PSA and safety indicators

Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

2004-11-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

Fifty years of Erlangen radiochemistry

International Nuclear Information System (INIS)

Morell, W.

2007-01-01

On June 29, 2006, the Radiochemical Laboratory of AREVA NP GmbH (formerly Siemens AG) in Erlangen celebrated its fiftieth anniversary. The occasion was marked by an event attended by more than 1,000 guests, among them Werner Gebauhr, the 85-year-old founder and first head of the Laboratory; the Managing Directors of AREVA NP GmbH, Ralf Gueldner and Ruediger Steuerlein; representatives of universities, research institutions, power utilities, and public authorities. The present head of the Radiochemical Laboratory, Wilfred Morell, sketched the highlights of the work performed over the past fifty years, which ranged from solid-state and very-high-purity materials technologies to development and service activities for nuclear technology. Manfred Erve, head of the Technical Center of AREVA NP GmbH, of which the Radiochemical Laboratory is a part, emphasized the changes in priorities over the past fifty years, which had always been met successfully by Radiochemistry. In the scientific part of the event, Wolfgang Schwarz (E.ON Kernkraftwerk GmbH, KKW Isar), Ulf Ilg (EnBW Kraftwerk AG, KKW Philippsburg), and Hans-Josef Allelein (Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH) explained 3 major subject areas in which Erlangen Radio-chemistry over many years has contributed basic findings (see other articles in this atw issue). On the occasion of the anniversary, a comprehensive booklet was published under the title of '50 Jahre Radiochemie Erlangen - 1956-2006'. (orig.)

Demonstration of inherent safety features of HTGRs using the HTTR

International Nuclear Information System (INIS)

Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

2004-01-01

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

International Nuclear Information System (INIS)

Charak, I.; Kier, P.H.

1995-04-01

Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

1997: Nuclear power plants in Europe. Report on operation, construction and planning in 18 European countries

International Nuclear Information System (INIS)

Anon.

1997-01-01

Eighteen European countries operate and build, respectively, nuclear power plants in 1997. The '1997: Nuclear Power Plants in Europe' atw report singles out the main events and lines of development. As per August 1997, 216 (1996: 216) nuclear generating units (which means power reactors for the purposes of this report) with an aggregate 180, 184 (177, 916) MWe installed gross capacity are in operation and 22 (24) units with 20,054 (23,086) MWe are under construction in seven countries. This adds up to a total of 238 (240) nuclear generating units with an aggregate 200,238 (201,002) MWe. In the nuclear power plants in Europe, some 1138 (1067) TWh of nuclear power was converted into electric power in 1996: 806 (802) TWh of this aggregate was converted in 142 (142) units in the European Union (EU). In the EU the share of nuclear power in the public supply of electricity was 35,4 (36,0)%. Lithuania, with 83,4 (85,6)% has the highest share of nuclear power in Europe, followed by France with 77,4 (76,0)% and Belgium with 57,2 (55,5)%. The lowest percentage, only 4,8 (3,8)%, is recorded in the Netherlands. As a consequence of electricity imports, nuclear power holds considerable shares in the public electricity supply also of countries in which no nuclear power is operated, such as Italy or Austria. (orig.) [de

Feedback of experience as a contribution to safety - an operator's duty

International Nuclear Information System (INIS)

Micklinghoff, M.; Pamme, H.; Raetzke, C.

2007-01-01

Exchanges of information about events and reliability problems make important contributions to nuclear power plant safety. In an article in the last but one issue of atw, journalist Timm Kraegenow proposed that the operators of nuclear power plants model their attitude towards this subject, and their flows of information, on the example of civil aviation. In particular, vendors should be made the nodes of information exchange. However, in-depth comparison shows that the two areas, civil aviation and nuclear power, have fundamentally different structures. The safety of aircraft designs continues to be the vendor's responsibility throughout the service life of that aircraft, and it is the vendor who holds the type certification. When a deficit becomes apparent, the vendor is the partner to be contacted by the competent authority, and it is the vendor's duty to elaborate solutions. In the case of nuclear power plants, however, responsibility for safety after plant commissioning rests with the operator, i.e. the licensee. The licensee initiates improvements in safety and is also the addressee of instructions by the authorities. This fundamental difference in responsibility for safety is in the nature of things. It is bound to affect also exchanges of information as a module of safety. Although comparison with civil aviation is interesting, it becomes apparent in the end that the way in which flows of information are designed cannot be transferred to the nuclear sector. (orig.)

RAMI analysis and modeling for the LANSCE accelerator systems

Energy Technology Data Exchange (ETDEWEB)

Macek, R.J.; Wilkinson, C.A. [Los Alamos National Laboratory, NM (United States)

1995-10-01

Reliability, availability, maintainability, and inspectability (RAMI) have become important issues for the high-power machines being planned for applications such as accelerator transmutation of nuclear waste (ATW), accelerator production of tritium (APT) and the next generation spallation neutron source. Beam reliability and beam availability are vitally important specifications to the present users of accelerator-driven spallation neutron sources, synchrotron light sources and medical accelerators. At Los Alamos, improved beam availability is a key goal in the planned LANSCE improvement program. Clearly, the capability to adequately model and predict the reliability and availability of complex accelerator systems will be of great value in assessing and optimizing RAMI measures in accelerator design and improvement programs. To date, no major accelerator project has developed comprehensive reliability models although the Advance Photon Source at ANL has started work on reliability analysis for selected subsystems. In this paper the authors discuss their experience in developing RAMI analysis and modeling for the LANSCE Accelerator Systems. Progress has been made in developing suitable measures and functions to characterize user risk, in logging of needed data on failure rates and repair/down times, and in developing a first-pass RAMI model for selected subsystems. Plans have been made for a more complete RAMI model. In addition, the authors discuss their experience in the use of probabilistic risk assessment (PRA) methodology for estimation of the reliability of active, instrumentation-based, radiation safety systems at LANSCE.

Development of One Meter Long Double-Sided CeO2 Buffered Ni-5at.%W Templates by Reel-to-Reel Chemical Solution Deposition Route

DEFF Research Database (Denmark)

Yue, Zhao; Konstantopoulou, K.; Wulff, Anders Christian

2013-01-01

High performance long-length coated conductors fabricated using various techniques have attracted a lot of interest recently. In this work, a reel-to-reel design for depositing double-sided coatings on long-length flexible metallic tapes via a chemical solution method is proposed and realized...... layer are 7.2◦ and 5.8◦ with standard deviation of 0.26◦ and 0.34◦, respectively, being indicative of the high quality epitaxial growth of the films prepared in the continuous manner. An all chemical solution derived YBCOLow−TFA/Ce0.9La0.1O2/Gd2Zr2O7/CeO2 structure is obtained on a short sample...

Surface engineering of biaxial Gd2Zr2O7 thin films deposited on Ni–5at%W substrates by a chemical solution method

DEFF Research Database (Denmark)

Yue, Zhao; Grivel, Jean-Claude; Liu, Min

2012-01-01

crystal structure along the film thickness observed by a transmission electron microscope. On the basis of the enhanced understanding of the crystallization processes, we demonstrate a possibility of engineering the surface morphology and texture in the film deposited on technical substrates using...... a chemical solution deposition route....

Transient and fuel performance analysis with VTT's coupled code system

International Nuclear Information System (INIS)

Daavittila, A.; Hamalainen, A.; Raty, H.

2005-01-01

VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

International Nuclear Information System (INIS)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

1994-01-01

The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

Simplified containment event tree analysis for the Sequoyah Ice Condenser containment

International Nuclear Information System (INIS)

Galyean, W.J.; Schroeder, J.A.; Pafford, D.J.

1990-12-01

An evaluation of a Pressurized Water Reactor (PER) ice condenser containment was performed. In this evaluation, simplified containment event trees (SCETs) were developed that utilized the vast storehouse of information generated by the NRC's Draft NUREG-1150 effort. Specifically, the computer programs and data files produced by the NUREG-1150 analysis of Sequoyah were used to electronically generate SCETs, as opposed to the NUREG-1150 accident progression event trees (APETs). This simplification was performed to allow graphic depiction of the SCETs in typical event tree format, which facilitates their understanding and use. SCETs were developed for five of the seven plant damage state groups (PDSGs) identified by the NUREG-1150 analyses, which includes: both short- and long-term station blackout sequences (SBOs), transients, loss-of-coolant accidents (LOCAs), and anticipated transient without scram (ATWS). Steam generator tube rupture (SGTR) and event-V PDSGs were not analyzed because of their containment bypass nature. After being benchmarked with the APETs, in terms of containment failure mode and risk, the SCETs were used to evaluate a number of potential containment modifications. The modifications were examined for their potential to mitigate or prevent containment failure from hydrogen burns or direct impingement on the containment by the core, (both factors identified as significant contributors to risk in the NUREG-1150 Sequoyah analysis). However, because of the relatively low baseline risk postulated for Sequoyah (i.e., 12 person-rems per reactor year), none of the potential modifications appear to be cost effective. 15 refs., 10 figs. , 17 tabs

Annual conference on nuclear technology. Nuclear power 2001: option for the future

International Nuclear Information System (INIS)

Anon.

2001-01-01

The Dresden Palace for Culture was the venue of the ANNUAL MEETING ON NUCLEAR TECHNOLOGY on May 15-17, 2001, the first to be held in Dresden and the first also to be held in one of the new German federal states. Although no nuclear plant is in operation in East Germany after the Greifswald Nuclear Power Station was decommissioned, nuclear technology continues to play an important role especially in research and university teaching in this part of Germany. The organizers of the conference, Deutsches Atomforum e.V. (DAtF) and Kerntechnische Gesellschaft e.V. (KTG), welcomed more than 1000 participants from nineteen countries. The three-day program, with its traditional, proven structure, featured plenary sessions on the first day, and specialized sessions, technical sessions, poster sessions, and other events on the following days. The partner country at the Annual Meeting on Nuclear Technology was Russia, with a session specially devoted to selected topics of the country. The conference was accompanied by a technical exhibition with company meeting points of vendors, suppliers, and service industries. A video film forum was arranged for the interested public which featured contributions about nuclear research, nuclear power plant operation, transport and storage as well as decommissioning. Another major event was a workshop on 'Preserving Competence in Nuclear Technology'. The plenary day is described in this summary report, while the results of the technical sessions as seen by the rapporteurs are printed elsewhere in this issue of atw 8/9, 2001. (orig.) [de

Fabrication of the cube textured NiO buffer layer by line-focused infrared heating for coated conductor application

International Nuclear Information System (INIS)

Chung, Jun-Ki; Kim, Won-Jeong; Tak, Jinsung; Kim, Cheol Jin

2007-01-01

Epitaxial growth of NiO on the bi-axially textured Ni-3 at.%W (Ni-3W) substrate as seed layer for coated conductor were studied. The bi-axially textured NiO was formed on the Ni-3W tapes using a line-focused infrared heater by oxidizing the surface of the substrate at 800-950 deg. C for 15-120 s in oxygen atmosphere. The thickness of the NiO layer could be controlled by changing heat-treatment, which was estimated as approximately 200-500 nm in the cross-sectional SEM micrographs of the NiO/Ni template. This thickness is enough to block the diffusion of the Ni in the substrate to the superconducting layer. The samples showed strong texture development of NiO layer. The sample oxidized at 900 deg. C with the tape transferring speed of 30 mm/h exhibited ω-scan full width at half maximum (FWHM) values for Ni-3W(2 0 0) and NiO(2 0 0) were 3.97 deg., and 3.67 deg., and φ-scan FWHM values for Ni-3W(1 1 1) and NiO(1 1 1) were 9.58 deg., and 8.79 deg., respectively. Also, the (1 1 1) pole-figure of the NiO buffer layer showed the good symmetry of the four peaks, securing the epitaxial growth of the buffer layers on the NiO layer. Also NiO layer exhibited root-mean-square roughness value of 39 nm by AFM (10 x 10 μm) investigation

RAMONA-4B development for SBWR safety studies

Energy Technology Data Exchange (ETDEWEB)

Rohatgi, U.S.; Aronson, A.L.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.

1993-12-31

The Simplified Boiling Water Reactor (SBWR) is a revolutionary design of a boiling-water reactor. The reactor is based on passive safety systems such as natural circulation, gravity flow, pressurized gas, and condensation. SBWR has no active systems, and the flow in the vessel is by natural circulation. There is a large chimney section above the core to provide a buoyancy head for natural circulation. The reactor can be shut down by either of four systems; namely, scram, Fine Motion Control Rod Drive (FMCRD), Alternate Rod Insertion (ARI), and Standby Liquid Control System (SLCS). The safety injection is by gravity drain from the Gravity Driven Cooling System (GDCS) and Suppression Pool (SP). The heat sink is through two types of heat exchangers submerged in the tank of water. These heat exchangers are the Isolation Condenser (IC) and the Passive Containment Cooling System (PCCS). The RAMONA-4B code has been developed to simulate the normal operation, reactivity transients, and to address the instability issues for SBWR. The code has a three-dimensional neutron kinetics coupled to multiple parallel-channel thermal-hydraulics. The two-phase thermal hydraulics is based on a nonhomogeneous nonequilibrium drift-flux formulation. It employs an explicit integration to solve all state equations (except for neutron kinetics) in order to predict the instability without numerical damping. The objective of this project is to develop a Sun SPARC and IBM RISC 6000 based RAMONA-4B code for applications to SBWR safety analyses, in particular for stability and ATWS studies.

Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

International Nuclear Information System (INIS)

Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

2000-01-01

The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

Evaluation of the PTS potential in a WWER-1000 following a steam line break

International Nuclear Information System (INIS)

Beghini, M.; D'Auria, F.; Galassi, G.M.; Vitale, E.

1997-01-01

A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs

Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Cuantificacion de secuencias de accidente y resultados

Energy Technology Data Exchange (ETDEWEB)

Huerta B, A; Aguilar T, O; Nunez C, A; Lopez M, R [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

1994-07-01

The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the {sup I}nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

RAMONA-4B development for SBWR safety studies

International Nuclear Information System (INIS)

Rohatgi, U.S.; Aronson, A.L.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.

1993-01-01

The Simplified Boiling Water Reactor (SBWR) is a revolutionary design of a boiling-water reactor. The reactor is based on passive safety systems such as natural circulation, gravity flow, pressurized gas, and condensation. SBWR has no active systems, and the flow in the vessel is by natural circulation. There is a large chimney section above the core to provide a buoyancy head for natural circulation. The reactor can be shut down by either of four systems; namely, scram, Fine Motion Control Rod Drive (FMCRD), Alternate Rod Insertion (ARI), and Standby Liquid Control System (SLCS). The safety injection is by gravity drain from the Gravity Driven Cooling System (GDCS) and Suppression Pool (SP). The heat sink is through two types of heat exchangers submerged in the tank of water. These heat exchangers are the Isolation Condenser (IC) and the Passive Containment Cooling System (PCCS). The RAMONA-4B code has been developed to simulate the normal operation, reactivity transients, and to address the instability issues for SBWR. The code has a three-dimensional neutron kinetics coupled to multiple parallel-channel thermal-hydraulics. The two-phase thermal hydraulics is based on a nonhomogeneous nonequilibrium drift-flux formulation. It employs an explicit integration to solve all state equations (except for neutron kinetics) in order to predict the instability without numerical damping. The objective of this project is to develop a Sun SPARC and IBM RISC 6000 based RAMONA-4B code for applications to SBWR safety analyses, in particular for stability and ATWS studies

Evaluation of the PTS potential in a WWER-1000 following a steam line break

Energy Technology Data Exchange (ETDEWEB)

Beghini, M; D` Auria, F; Galassi, G M; Vitale, E [Universita degli Studi di Pisa, Dipt. di Costruzioni Meccaniche e Nucleari, Pisa (Italy)

1997-09-01

A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs.

Uncertainty assessment for accelerator-driven systems

International Nuclear Information System (INIS)

Finck, P. J.; Gomes, I.; Micklich, B.; Palmiotti, G.

1999-01-01

The concept of a subcritical system driven by an external source of neutrons provided by an accelerator ADS (Accelerator Driver System) has been recently revived and is becoming more popular in the world technical community with active programs in Europe, Russia, Japan, and the U.S. A general consensus has been reached in adopting for the subcritical component a fast spectrum liquid metal cooled configuration. Both a lead-bismuth eutectic, sodium and gas are being considered as a coolant; each has advantages and disadvantages. The major expected advantage is that subcriticality avoids reactivity induced transients. The potentially large subcriticality margin also should allow for the introduction of very significant quantities of waste products (minor Actinides and Fission Products) which negatively impact the safety characteristics of standard cores. In the U.S. these arguments are the basis for the development of the Accelerator Transmutation of Waste (ATW), which has significant potential in reducing nuclear waste levels. Up to now, neutronic calculations have not attached uncertainties on the values of the main nuclear integral parameters that characterize the system. Many of these parameters (e.g., degree of subcriticality) are crucial to demonstrate the validity and feasibility of this concept. In this paper we will consider uncertainties related to nuclear data only. The present knowledge of the cross sections of many isotopes that are not usually utilized in existing reactors (like Bi, Pb-207, Pb-208, and also Minor Actinides and Fission Products) suggests that uncertainties in the integral parameters will be significantly larger than for conventional reactor systems, and this raises concerns on the neutronic performance of those systems

Prediction of Fission Product Release during the LOFC Experiments at the HTTR

International Nuclear Information System (INIS)

Shi, D.; Xhonneux, A.; Verfondern, K.; Ueta, S.; Allelein, H.-J.

2014-01-01

Demonstration tests were conducted using the High Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, to confirm the safety of HTGR technologies and assure the expected physical phenomena to occur under given conditions. As part of the OECD directed LOFC (“loss of forced cooling”) project, a series of three tests at the HTTR has been planned with tripping of all gas circulators while deactivating all reactor reactivity control to disallow reactor scram due to abnormal reduction of primary coolant flow rate. The tests fall into anticipated transient without scram (ATWS) with occurrence of reactor recriticality. They serve the important purpose to provide a valuable data base for the validation of computer models regarding neutronics, heat transfer and fluid dynamics, fuel performance and fission product transport and release behavior in HTGRs. The Source Term Analysis Code System (STACY) is a new code development at the Research Center Jülich encompassing the original verified and validated computer models for simulating fission product transport and release. For verification of the modernized and extended version, it was assured that results obtained with the original tools could be reproduced. One of the new features of STACY is its ability to also treat fuel compacts of (full) cylindrical or annular shape and a complete prismatic block reactor core, respectively, supposed sufficient input data be available. The paper will describe the new STACY tool and present the results of fission product behavior in the HTTR core under the LOFC test conditions. Calculations are based on time-dependent neutronics and fluid dynamics results obtained with the Serpent and MGT models. (author)

Uncertainty analysis for hot channel

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2006-01-01

The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

University Programs of the U.S. Department of Energy Advanced Accelerator Applications Program

International Nuclear Information System (INIS)

Beller, Denis E.; Ward, Thomas E.; Bresee, James C.

2002-01-01

The Advanced Accelerator Applications (AAA) Program was initiated in fiscal year 2001 (FY-01) by the U.S. Congress, the U.S. Department of Energy (DOE), and the Los Alamos National Laboratory (LANL) in partnership with other national laboratories. The primary goal of this program is to investigate the feasibility of transmutation of nuclear waste. An Accelerator-Driven Test Facility (ADTF), which may be built during the first decade of the 21. Century, is a major component of this effort. The ADTF would include a large, state-of-the-art charged-particle accelerator, proton-neutron target systems, and accelerator-driven R and D systems. This new facility and its underlying science and technology will require a large cadre of educated scientists and trained technicians. In addition, other applications of nuclear science and engineering (e.g., proliferation monitoring and defense, nuclear medicine, safety regulation, industrial processes, and many others) require increased academic and national infrastructure and student populations. Thus, the AAA Program Office has begun a multi-year program to involve university faculty and students in various phases of the Project to support the infrastructure requirements of nuclear energy, science and technology fields as well as the special needs of the DOE transmutation program. In this paper we describe university programs that have supported, are supporting, and will support the R and D necessary for the AAA Project. Previous work included research for the Accelerator Transmutation of Waste (ATW) project, current (FY-01) programs include graduate fellowships and research for the AAA Project, and it is expected that future programs will expand and add to the existing programs. (authors)

Correlates of volunteering among aging Texans: the roles of health indicators, spirituality, and social engagement.

Science.gov (United States)

Ahn, SangNam; Phillips, Karon L; Smith, Matthew Lee; Ory, Marcia G

2011-07-01

This study aimed to identify participant characteristics associated with volunteering among older adults. Based on data from the 2008 Aging Texas Well (ATW) Indicators Survey, we examined the degree to which demographic factors, health status, spiritual participation, and community involvement are associated with volunteering among adults aged 60 years or older (n = 525). Rates of volunteering varied by race/ethnicity: non-Hispanic Whites (56.4%), African Americans (51.1%), and Hispanics (43.2%). Bivariate analyses showed that non-Hispanic White older adults were more likely to participate in formal volunteering activities, while their African American and Hispanic counterparts tended to participate in informal volunteering activities. Logistic regression analyses revealed that volunteering was less observed among Hispanics (OR = 0.48, 95% CI 0.29-0.78). Volunteering was more observed among those who reported providing informal care (OR=1.93, 95% CI 1.14-3.28), having very good or excellent mental health (OR = 1.90 and 2.07, 95% CI 1.09-3.32 and 1.20-3.55, respectively), having weekly or daily spiritual participation (OR = 2.15 and 2.35, 95% CI 1.28-3.63 and 1.29-4.28, respectively), perceiving community involvement very important (OR = 2.37, 95% CI 1.55-3.62), and being very satisfied with the community interaction (OR = 1.81, 95% CI 1.15-2.85). Given the positive associations of mental health, spirituality, and social engagement with volunteering among older adults, system-level efforts to increase the sense of community among older adults and recognize their roles as volunteers will be helpful in recruiting and retaining older volunteers. Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

YKAe Research programme on nuclear power plant systems behaviour and operational aspects of safety 1990-1994, Final report

International Nuclear Information System (INIS)

Mattila, L.; Vanttola, T.

1995-04-01

The research programme on Nuclear Power Plant Systems Behaviour and Operational Aspects of Safety was carried out between 1990 and 1994. In the field of Safe operational margins of nuclear fuel and reactor core, an up-to-date steady-state fuel performance model was validated for higher burn-ups and well-characterized VVER fuel experiments were carried out. A comprehensive reactor analysis code system was extended and validated for complex 3-D phenomena, such as ATWS and boron dilution transients. Advanced hydraulics methods were added to the dynamics codes. Experiments were carried out with PACTEL, the most comprehensive thermal-hydraulic test facility for VVER-440-type reactors worldwide. For example, a series of natural circulation tests were provided for the first VVER-related international standard problem of the OECD/NEA. Advanced foreign computer codes for severe accidents were evaluated and modified for the needs of Finnish power plants. Specific progress was made in modelling the reflooding of an overheated core and in the structural analysis of a pressure vessel under corium load, as well as in experimental and theoretical studies of aerosol and hydrogen behaviour. Fire modelling was improved by implementing advanced calculation methods and by validating them against our own experiments and international test data. Techniques in living probabilistic safety assessment and risk decision-making were refined in pilot applications for continuous monitoring, follow-up and management of risks of an operating power plant. In the area of human reliability and organizational performance, factors important for the continuous development of the competence of control room operator teams and plant maintenance staff were identified. (237 refs., 75 figs., 13 tabs.)

Comparative Performance with Different Versions of Low Heat Rejection Combustion Chambers with Crude Rice Bran Oil

Directory of Open Access Journals (Sweden)

Krishna M.V.S. Murali

2014-12-01

Full Text Available Jak wiadomo, oleje roslinne sa obiecujacym substytutem paliw ropopochodnych, poniewaz ich własciwosci sa podobne do oleju dieslowskiego, sa odnawialne i łatwe do wyprodukowania. Niemniej, surowe oleje roslinne wykazuja wady, takie jak wysoka lepkosc i mała lotnosc, co wymaga komory spalania o małych stratach ciepła, której istotnymi cechami sa m.in. wyzsza temperatura robocza, maksymalne wydzielanie ciepła i zdolnosc do wykorzystania paliwa o mniejszej wartosci kalorycznej (CV. Przeprowadzono eksperymenty majace na celu ocene osiagów silnika z róznymi komorami spalania o małych stratach ciepła (LHR, takich jak głowica cylindra o pokryciu ceramicznym (LHR-1, tłok izolowany szczelina powietrzna z denkiem ze stopu Superni (superstop niklu i tuleja cylindra z wkładka z Superni izolowana szczelina powietrzna (LHR-2 oraz głowica cylindra z pokryciem ceramicznym, tłok i tuleja cylindra izolowane szczelinami powietrznymi (LHR-3. Badania prowadzono przy normalnej temperaturze oleju roslinnego (surowy olej z otrab ryzowych, CRBO i zmiennym cisnieniu w otworze wtryskiwacza. Parametry osiagów silnika (uzyteczna sprawnosc termiczna, uzyteczny współczynnik zuzycia energii, temperatura gazu wydychanego, obciazenie obiegiem chłodziwa i współczynnik napełnienia oraz emisje wydechowe [poziomy dymu i tlenków azotu, NOx] zostały wyznaczone przy róznych wartosciach sredniego uzytecznego cisnienia w silniku. Charakterystyki spalania [cisnienie szczytowe, czas wystepowania cisnienia szczytowego, maksymalna szybkosc wzrostu cisnienia] zostały wyznaczone w warunkach pracy silnika z pełnym obciazeniem.

Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

International Nuclear Information System (INIS)

Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

2010-02-01

A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

Textured YBCO films grown on wires: application to superconducting cables

International Nuclear Information System (INIS)

Dechoux, N; Jiménez, C; Chaudouët, P; Rapenne, L; Sarigiannidou, E; Robaut, F; Petit, S; Garaudée, S; Porcar, L; Soubeyroux, J L; Odier, P; Bruzek, C E; Decroux, M

2012-01-01

Efforts to fabricate superconducting wires made of YBa 2 Cu 3 O 7 (YBCO) on La 2 Zr 2 O 7 (LZO) buffered and biaxially textured Ni-5 at.%W (NiW) are described. Wires were manually shaped from LZO buffered NiW tapes. Different diameters were produced: 1.5, 2 and 3 mm. The wires were further covered with YBCO grown by metal organic chemical vapor deposition (MOCVD). We developed an original device in which the round substrate undergoes an alternated rotation of 180° around its axis in addition to a reel-to-reel translation. This new approach allows covering the whole circumference of the wire with a YBCO layer. This was confirmed by energy dispersive x-ray spectroscopy (EDX) analysis coupled to a scanning electron microscope (SEM). For all wire diameters, the YBCO layer thickness varied from 300 to 450 nm, and the cationic composition was respected. Electron backscattering diffraction (EBSD) measurements were performed directly on an as-deposited wire without surface preparation allowing the investigation of the crystalline quality of the film surface. Combining EBSD with XRD results we show that YBCO grows epitaxially on the LZO buffered NiW wires. For the first time, superconductive behaviors have been detected on round substrates in both the rolling and circular direction. J c reached 0.3 MA cm −2 as measured at 77 K by transport and third-harmonic detection. Those preliminary results confirm the effectiveness of the MOCVD for complex geometries, especially for YBCO deposition on small diameter wires. This approach opens huge perspectives for the elaboration of a new generation of YBCO-based round conductors. (paper)

Technologies for destruction of long-lived radionuclides in high-level nuclear waste - overview and requirements

International Nuclear Information System (INIS)

Arthur, E.D.

1993-01-01

A major issue surrounding current nuclear power generation is the management and disposal of long-lived, high-level waste (HLW). The planned and scientifically acceptable destination for this waste is in deep underground, geologically stable, repositories. However, public concerns surrounding such disposal of long-lived nuclear wastes and other issues such as proliferation and safety negatively affect the potential role that nuclear power can play in meeting current and future national energy needs. This paper and this topical session on nuclear waste minimization, management, and remediation focus on two nuclear systems and their associated technologies that have the potential to address concerns surrounding long-lived radionuclides in HLW. Both systems offer technology applicable to HLW from current light water reactors (LWRs). In addition, these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes long-term fuel supplies, and enhanced safety. The first system is the integral fast reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while burning long-lived actinide waste. The IFR applications include the burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, accelerator transmutation of waste (ATW), is based on an accelerator-induced intense source of thermal neutrons and is aimed at the destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent-fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

International Nuclear Information System (INIS)

Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

2010-07-01

Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station

International Nuclear Information System (INIS)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

1993-01-01

This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

Development of regulatory technology for thermal-hydraulic safety analysis

International Nuclear Information System (INIS)

Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

2001-02-01

The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

Neutron and proton transmutation-activation cross section libraries to 150 MeV for application in accelerator-driven systems and radioactive ion beam target-design studies

International Nuclear Information System (INIS)

Koning, A.J.; Chadwick, M.B.; MacFarlane, R.E.; Mashnik, S.; Wilson, W.B.

1998-05-01

New transmutation-activation nuclear data libraries for neutrons and protons up to 150 MeV have been created. These data are important for simulation calculations of radioactivity, and transmutation, in accelerator-driven systems such as the production of tritium (APT) and the transmutation of waste (ATW). They can also be used to obtain cross section predictions for the production of proton-rich isotopes in (p,xn) reactions, for radioactive ion beam (RIB) target-design studies. The nuclear data in these libraries stem from two sources: for neutrons below 20 MeV, we use data from the European activation and transmutation file, EAF97; For neutrons above 20 MeV and for protons at all energies we have isotope production cross sections with the nuclear model code HMS-ALICE. This code applies the Monte Carlo Hybrid Simulation theory, and the Weisskopf-Ewing theory, to calculate cross sections. In a few cases, the HMS-ALICE results were replaced by those calculated using the GNASH code for the Los Alamos LA150 transport library. The resulting two libraries, AF150.N and AF150.P, consist of 766 nuclides each and are represented in the ENDF6-format. An outline is given of the new representation of the data. The libraries have been checked with ENDF6 preprocessing tools and have been processed with NJOY into libraries for the Los Alamos transmutation/radioactivity code CINDER. Numerous benchmark figures are presented for proton-induced excitation functions of various isotopes compared with measurements. Such comparisons are useful for validation purposes, and for assessing the accuracy of the evaluated data. These evaluated libraries are available on the WWW at: http://t2.lanl.gov/. 21 refs

ESBWR - Robust design for natural circulation and stability performance effectiveness

Energy Technology Data Exchange (ETDEWEB)

Alamgir, M. D.; Marquino, W.; Yang, J.; Saha, P.; Fennern, L.; Colby, M. [GE-Hitachi Nuclear Energy, M/C A65, 3901Castle Hayne Road, Wilmington, NC 28401 (United States)

2012-07-01

ESBWR is a 4500 MWt Generation III+ natural circulation reactor with an array of robust design features and passive safety systems to deliver highly effective plant performance during normal operation and to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US design certification process. This paper focuses on the natural circulation-driven plant performance aspects during normal operation, and stability evaluation of the robust ESBWR design. The TRACG computer code is used for the analysis of ESBWR plant performance, safety analysis, and stability margins. The paper describes the evaluation of ESBWR stability performance during normal power operation including operation in the Core Power-Feed Water Temperature Operating Domain. For ESBWR the normal power operation condition has the highest power/flow ratio and is limiting from the perspective of stability. The paper includes results from detailed evaluation of the most limiting decay ratio for out-of-phase regional oscillations calculated by perturbing the core inlet flow rate in this out-of-phase mode about the line of symmetry for the azimuthal harmonic mode. The paper also summarizes the ESBWR regional mode stability evaluations during a limiting transient (Loss of Feedwater Heating), and during ATWS (Anticipated Transient without Scram). Nominal decay ratios of limiting Channel oscillation, Core wide oscillation and Regional oscillation are within the maximum acceptance criterion of 0.8, at 95% content and 95% confidence. These stability evaluation results indicate decay ratio is within design limits. The paper also describes the evaluation of ESBWR stability performance during plant startup, and summarizes the defense-in-depth stability solution for ESBWR. (authors)

ESBWR - Robust design for natural circulation and stability performance effectiveness

International Nuclear Information System (INIS)

Alamgir, M. D.; Marquino, W.; Yang, J.; Saha, P.; Fennern, L.; Colby, M.

2012-01-01

ESBWR is a 4500 MWt Generation III+ natural circulation reactor with an array of robust design features and passive safety systems to deliver highly effective plant performance during normal operation and to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US design certification process. This paper focuses on the natural circulation-driven plant performance aspects during normal operation, and stability evaluation of the robust ESBWR design. The TRACG computer code is used for the analysis of ESBWR plant performance, safety analysis, and stability margins. The paper describes the evaluation of ESBWR stability performance during normal power operation including operation in the Core Power-Feed Water Temperature Operating Domain. For ESBWR the normal power operation condition has the highest power/flow ratio and is limiting from the perspective of stability. The paper includes results from detailed evaluation of the most limiting decay ratio for out-of-phase regional oscillations calculated by perturbing the core inlet flow rate in this out-of-phase mode about the line of symmetry for the azimuthal harmonic mode. The paper also summarizes the ESBWR regional mode stability evaluations during a limiting transient (Loss of Feedwater Heating), and during ATWS (Anticipated Transient without Scram). Nominal decay ratios of limiting Channel oscillation, Core wide oscillation and Regional oscillation are within the maximum acceptance criterion of 0.8, at 95% content and 95% confidence. These stability evaluation results indicate decay ratio is within design limits. The paper also describes the evaluation of ESBWR stability performance during plant startup, and summarizes the defense-in-depth stability solution for ESBWR. (authors)

Aspiration tests in aqueous foam using a breathing simulator

Energy Technology Data Exchange (ETDEWEB)

Archuleta, M.M.

1995-12-01

Non-toxic aqueous foams are being developed by Sandia National Laboratories (SNL) for the National Institute of Justice (NIJ) for use in crowd control, cell extractions, and group disturbances in the criminal justice prison systems. The potential for aspiration of aqueous foam during its use and the resulting adverse effects associated with complete immersion in aqueous foam is of major concern to the NIJ when examining the effectiveness and safety of using this technology as a Less-Than-Lethal weapon. This preliminary study was designed to evaluate the maximum quantity of foam that might be aspirated by an individual following total immersion in an SNL-developed aqueous foam. A.T.W. Reed Breathing simulator equipped with a 622 Silverman cam was used to simulate the aspiration of an ammonium laureth sulfate aqueous foam developed by SNL and generated at expansion ratios in the range of 500:1 to 1000:1. Although the natural instinct of an individual immersed in foam is to cover their nose and mouth with a hand or cloth, thus breaking the bubbles and decreasing the potential for aspiration, this study was performed to examine a worst case scenario where mouth breathing only was examined, and no attempt was made to block foam entry into the breathing port. Two breathing rates were examined: one that simulated a sedentary individual with a mean breathing rate of 6.27 breaths/minute, and one that simulated an agitated or heavily breathing individual with a mean breathing rate of 23.7 breaths/minute. The results of this study indicate that, if breathing in aqueous foam without movement, an air pocket forms around the nose and mouth within one minute of immersion.

TASS code topical report. V.1 TASS code technical manual

International Nuclear Information System (INIS)

Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

1997-02-01

TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

Summary of session W2

International Nuclear Information System (INIS)

Deitrich, L.W.

1988-01-01

This paper reports on the development and validation of the SASSYS-1 code which is being used to analyze the transient behavior of the U.S. innovative designs, SAFR and PRISM, as well as the future metal fueled core for FFTF. The description emphasized the high degree of flexibility of SASSYS-1 and its capability to do detailed core calculations needed to accurately model reactivity feedbacks. Validation work based on calculations of EBR-II SHRT (shutdown neat removal) tests and FFTF transient experiments was presented. The paper presents a discussion of the modifications to the SSC code required to model the ATWS events in SAFR and PRISM, along with results of calculations done with the modified SSC. The probability that passive shutdown could be defeated and severe core damage sustained has been investigated, and some results are presented in this paper. It was stated that simple design choices can keep risk almost arbitrarily low, and that the leading uncertainties will be reduced through future R ampersand D and demonstration testing. Uncertainties exist due to uncertainties in feedback coefficients and in the models and codes used to do the calculations. Large margin and relatively small uncertainties lead to low probabilities (less than 10-3) of defeat of passive shutdown by exceeding temperature limits. The given value is considered to be conservative to allow for uncertainties in knowledge of all relevant accident sequences at this stage in design and analysis. This paper also deals with experimental work on decay heat removal in PRISM and SAFR. Experiments on air-side heat transfer in the RVACS/RACS passive decay heat removal path, was reported. A data correlation for use in predicting heat transfer has been discussed for the RVACS conditions

Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

1993-07-01

This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

International Nuclear Information System (INIS)

Beare, A.N.; Dorris, R.E.; Gray, L.H.

1984-12-01

This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

Research on CDA for advanced fuel FBR

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime.

1997-03-01

For the purpose of evaluating possibility of the re-criticality of a metallic fueled reactor, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'Research on CDA for advanced fuel FBR'. The results of this year are the following. The accident initiator considered is a loss-of-flow accident with ATWS. The LOF analysis was performed for the metallic fueled 600 MWe homogeneous two region reactors, both for a metallic fuel only and for a metallic fuel core with ZrH pin. The SAS3D CDA initiation phase analysis code was used to investigate the re-criticality potential at the severe accident. The change mainly in the constants was necessary to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. LOF with flow decay half time of t 1/2 =0.5(s) (all blackout case) and 5.5(s) (ordinary LOF case) were analyzed. Independent of the conditions of the analysis, the results show all the cases could avoid to become prompt-critical. Depending on the analysis condition, it becomes necessary to transfer to the transient phase, it is also shown there is a possibility to avoid re-criticality due to the motion of molten fuel both for the metallic fuel and for the metallic fuel with ZrH moderator. However, because of the constants used for the material property the results might overestimate the fuel motion. It is shown that the moderator is effective to terminate the accident at an early stage. The behavior of metallic fueled reactors at CDA was analyzed with SAS3D code by modifying the constants of material properties to be applied to the reactor. It is shown that a metallic fueled reactor has a possibility to avoid re-criticality at CDA. (J.P.N.)

Life extension of German nuclear power plants only with the consent of the Federal Council? The importance and extent of the need for consent to an amendment to the German Atomic Energy Act

International Nuclear Information System (INIS)

Schneider, Horst

2010-01-01

In its coalition agreement of October 26, 2009, the new German federal government plans ''to extend the service life of German nuclear power plants while, at the same time, complying with the strict German and international safety standards.'' This has triggered a debate not only about (nuclear) energy, as in the past election campaign in the summer of 2009, but also about the constitutional law issue whether an amendment to the Atomic Energy Act resulting in longer operating life of nuclear power plants required the consent of the Federal Council (the ''Bundesrat,'' the second chamber of parliament). After the election to the state parliament in North Rhine-Westphalia on May 9, 2010, majority in the Federal Council changed. As a consequence, no consent to an amendment to the Atomic Energy Act must be expected. In view of the large number of recent statements about constitutional law in opinions for various federal and ministerial accounts as well as firms and associations, the outline by R. Scholz in the May issue of atw 2010 will be followed in this issue by the key points of examination of the need for consent, under aspects of constitutional law, and an attempt will be made to explain the evaluations underlying the generation of a legal concept about these items. The decision by the German Federal Constitutional Court of May 4, 2010, published on June 11, 2010, plays a major role in this respect because it established clarity in some important aspects of a legal subject matter in the field of state admini-stration on behalf of the federation, albeit in the field of air traffic law, not nuclear law. However, the structures of the norms in the German Basic Law (Art. 87c and Art. 87d, para.2) to be applied are almost identical. The energy policy and energy economy aspects of a plant life extension are considered along with the option of an appeal to the Federal Constitutional Court against any plant life extension. Finally, the key findings are summarized briefly

Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

International Nuclear Information System (INIS)

Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook

2007-08-01

This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the modeling

Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

International Nuclear Information System (INIS)

Papukchiev, Angel; Schaefer, Anselm

2008-01-01

In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

Numerical simulation of boron injection in a BWR

Energy Technology Data Exchange (ETDEWEB)

Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

2010-02-15

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

Numerical simulation of boron injection in a BWR

International Nuclear Information System (INIS)

Tinoco, Hernan; Buchwald, Przemyslaw; Frid, Wiktor

2010-01-01

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several

Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

2007-08-15

This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

Comparative analysis of sub-critical transmutation reactor concepts

International Nuclear Information System (INIS)

Chang, S. H.

1997-01-01

The long-lived nuclear wastes have been substantially generated from the light water reactor for a few decades. The toxicity of these spent fuels will be higher than that of the uranium ore, even if those will be stored in the repository more than ten thousands. Hence the means of transmuting the key long-lived nuclear wastes, primarily the minor actinides, using a hybrid proton accelerator and subcritical transmutation reactor, are proposed. Until now, the representative concepts for a subcritical transmutation reactor are the Energy Amplifier, the OMEGA project, the ATW and the MSBR. The detailed concepts and the specifications are illustrated in Table 1. The design requirements for the subcritical transmutation reactor are the high transmutation rate of long-lived nuclear wastes, safety and economics. And to propose the subcritical transmutation reactor concepts, the coolant, the target material and fuel type are carefully considered. In these aspects, the representative concepts for a subcritical transmutation reactor in Table 1 have been surveyed. The requirements for a target and a coolant are the reliable, low maintenance operation and safe operation to minimize the wastes. The reliable, low maintenance operation and safe operation to minimize the wastes. The reliable coolant must have the low melting point, high heat capacity and excellent physical properties. And the target material must have high neutron yield for a given proton condition and easy heat removal capability. Therefore in respect with the above requirements, Pb-Bi is proposed as the coolant and the target material for the subcritical reactor. Because the neutron yield for a given proton energy increases linearly with mass number up to bismuth but in heavier elements spallation events sharply increase both the neutron and heat outputs, Pb-Bi meets not only such the requirements as the above for the coolant but also those for the coolant and target, the simplification of system can be achieved

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

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; 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

Numerical Modeling of Lead Oxidation in Controlled Lead Bismuth Eutectic Systems: Chemical Kinetics and Hydrodynamic Effects

International Nuclear Information System (INIS)

Wu, Chao; Kanthi Kiran Dasika; Chen, Yitung; Moujaes, Samir

2002-01-01

Using liquid Lead-Bismuth Eutectic (LBE) as coolant in nuclear systems has been studied for more than 50 years. And LBE has many unique nuclear, thermo physical and chemical attributes which are attractive for practical application. But, corrosion is one of the greatest concerns in using liquid Lead-Bismuth Eutectic (LBE) as spallation target in the Accelerator-driven Transmutation of Waste (ATW) program. Los Alamos National Laboratory has designed and built the Liquid Lead-Bismuth Materials Test Loop (MTL) to study the materials behavior in a flow of molten LBE. A difference of 100 deg. C was designed between the coldest and the hottest parts at a nominal flow rate of 8.84 GPM. Liquid LBE flow was activated by a mechanical sump pump or by natural convection. In order to maintain a self-healing protective film on the surface of the stainless steel pipe, a certain concentration of oxygen has to be maintained in the liquid metal. Therefore, it is of importance to understand what the oxygen concentrations are in the LBE loop related to the corrosion effects on the metal surface, the temperature profiles, the flow rates, and diffusion rates through the metal surface. The chemical kinetics also needs to be fully understood in the corrosion processes coupled with the hydrodynamics. The numerical simulation will be developed and used to analyze the system corrosion effects with different kind of oxygen concentrations, flow rates, chemical kinetics, and geometries. The hydrodynamics modeling of using computational fluid dynamics will provide the necessary the levels of oxygen and corrosion products close to the boundary or surface. This paper presents an approach towards the above explained tasks by analyzing the reactions between the Lead and oxygen at a couple of sections in the MTL. Attempt is also made to understand the surface chemistry by choosing an example model and estimating the near wall surface concentration values for propane and oxygen. (authors)

The significance of biometric parameters in determining anterior teeth width

Directory of Open Access Journals (Sweden)

Strajnić Ljiljana

2013-01-01

Full Text Available Background/Aim. An important element of prosthetic treatment of edentulous patients is selecting the size of anterior artificial teeth that will restore the natural harmony of one’s dentolabial structure as well as the whole face. The main objective of this study was to determine the correlation between the inner canthal distance (ICD and interalar width (IAW on one side and the width of both central incisors (CIW, the width of central and lateral incisors (CLIW, the width of anterior teeth (ATW, the width between the canine cusps (CCW, which may be useful in clinical practice. Methods. A total of 89 subjects comprising 23 male and 66 female were studied. Their age ranged from 19 to 34 years with the mean of 25 years. Only the subjects with the preserved natural dentition were included in the sample. All facial and intraoral tooth measurements were made with a Boley Gauge (Buffalo Dental Manufacturing Co., Brooklyn NY, USA having a resolution of 0.1mm. Results. A moderate correlation was established between the interalar width and combined width of anterior teeth and canine cusp width (r = 0.439, r = 0.374. A low correlation was established between the inner canthal distance and the width of anterior teeth and canine cusp width (r = 0.335, r = 0.303. The differences between the two genders were highly significant for all the parameters (p < 0.01. The measured facial distances and width of anterior teeth were higher in men than in women. Conclusion. The results of this study suggest that the examined interalar width and inner canthal distance cannot be considered reliable guidelines in the selection of artificial upper anterior teeth. However, they may be used as a useful additional factor combined with other methods for objective tooth selection. The final decision should be made while working on dentures fitting models with the patient’s consent.

Implementing standards for the interoperability among healthcare providers in the public regionalized Healthcare Information System of the Lombardy Region.

Science.gov (United States)

Barbarito, Fulvio; Pinciroli, Francesco; Mason, John; Marceglia, Sara; Mazzola, Luca; Bonacina, Stefano

2012-08-01

Information technologies (ITs) have now entered the everyday workflow in a variety of healthcare providers with a certain degree of independence. This independence may be the cause of difficulty in interoperability between information systems and it can be overcome through the implementation and adoption of standards. Here we present the case of the Lombardy Region, in Italy, that has been able, in the last 10 years, to set up the Regional Social and Healthcare Information System, connecting all the healthcare providers within the region, and providing full access to clinical and health-related documents independently from the healthcare organization that generated the document itself. This goal, in a region with almost 10 millions citizens, was achieved through a twofold approach: first, the political and operative push towards the adoption of the Health Level 7 (HL7) standard within single hospitals and, second, providing a technological infrastructure for data sharing based on interoperability specifications recognized at the regional level for messages transmitted from healthcare providers to the central domain. The adoption of such regional interoperability specifications enabled the communication among heterogeneous systems placed in different hospitals in Lombardy. Integrating the Healthcare Enterprise (IHE) integration profiles which refer to HL7 standards are adopted within hospitals for message exchange and for the definition of integration scenarios. The IHE patient administration management (PAM) profile with its different workflows is adopted for patient management, whereas the Scheduled Workflow (SWF), the Laboratory Testing Workflow (LTW), and the Ambulatory Testing Workflow (ATW) are adopted for order management. At present, the system manages 4,700,000 pharmacological e-prescriptions, and 1,700,000 e-prescriptions for laboratory exams per month. It produces, monthly, 490,000 laboratory medical reports, 180,000 radiology medical reports, 180

Recent developments in nuclear data for ADS

International Nuclear Information System (INIS)

Talou, P.; Chadwick, M.B.; Young, P.G.

2003-01-01

systematic uncertainties which always affect experimental results. We will present the Bayesian inference scheme as a powerful tool to achieve this goal, and will illustrate this concept with the example of the evaluation of 239 Pu (n,f), recently assessed in the context of the ATW/AAA [Accelerator Transmutation of Waste / Advanced Accelerator Applications] program in Los Alamos. (author)

Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

Energy Technology Data Exchange (ETDEWEB)

Chen, Che-Hao; Shih, Chunkuan [National Tsing Hua Univ., Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Lin, Hao-Tzu [Atomic Energy Council, Taiwan (China). Inst. of Nuclear Energy Research

2013-07-01

Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

International Nuclear Information System (INIS)

Chen, Che-Hao; Shih, Chunkuan; Wang, Jong-Rong; Lin, Hao-Tzu

2013-01-01

Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

Ból w klatce piersiowej – podobne objawy, a różne rozpoznania i przebieg kliniczny – opisy przypadków

Directory of Open Access Journals (Sweden)

Anna Kaźmierczak-Dziuk

2011-12-01

Full Text Available Ból w klatce piersiowej jest najczęstszym objawem chorób układu krążenia, może być jednak spowodowany zaburzeniami ze strony innych układów i narządów. Ustalenie przyczyny bólu nie jest łatwe, niemniej jednak kluczowe w procesie diagnostycznym. W różnicowaniu należy wziąć pod uwagę przede wszystkim choroby najczęściej występujące oraz te, które zagrażają bezpośrednio życiu i zdrowiu pacjenta. Są to stany związane z niedokrwieniem mięśnia serca: niestabilna dławica piersiowa, zawał mięśnia serca. Podczas oceny chorego konieczne jest uwzględnienie przyczyn bólu w klatce piersiowej pochodzących z wnętrza klatki piersiowej (aorta, tętnica płucna, śródpiersie, przełyk, tchawica, oskrzela, opłucna, ze ściany klatki piersiowej (nerwy czuciowe, połączenia chrzęstno-kostne, kręgosłup, skóra, ze strony jamy brzusznej (żołądek, dwunastnica, trzustka, pęcherzyk żółciowy, a także dolegliwości o podłożu psychogennym. W procesie diagnostycznym należy brać pod uwagę charakter bólu, jego czas trwania, sytuacje prowokujące i zmniejszające ból, a także objawy towarzyszące. Do czasu zakończenia postępowania diagnostycznego chory z bólem w klatce piersiowej powinien być traktowany jak pacjent o istotnie podwyższonym ryzyku sercowo-naczyniowym. Dokładnie przeprowadzona diagnostyka różnicowa pozwala na ustalenie prawidłowego rozpoznania i włączenie odpowiedniego leczenia. W pracy przedstawiono opisy pięciu przypadków klinicznych chorych z bólem w klatce piersiowej. Pomimo podobnych objawów u każdego z prezentowanych pacjentów ustalono inne rozpoznanie, a przebieg kliniczny był w każdym przypadku zupełnie odmienny.

Recent developments in nuclear data for ADS

Energy Technology Data Exchange (ETDEWEB)

Talou, P [T-16, Nuclear Physics Group, Theoretical Division, Los Alamos National Laboratory, Los Alamos (United States); Chadwick, M B [T-16, Nuclear Physics Group, Theoretical Division, Los Alamos National Laboratory, Los Alamos (United States); Young, P G [T-16, Nuclear Physics Group, Theoretical Division, Los Alamos National Laboratory, Los Alamos (United States)

2003-06-15

systematic uncertainties which always affect experimental results. We will present the Bayesian inference scheme as a powerful tool to achieve this goal, and will illustrate this concept with the example of the evaluation of {sup 239}Pu (n,f), recently assessed in the context of the ATW/AAA [Accelerator Transmutation of Waste / Advanced Accelerator Applications] program in Los Alamos. (author)

Results from Accelerator Driven TRIGA Reactor Experiments at The University of Texas at Austin

International Nuclear Information System (INIS)

O'Kelly, S.; Braisted, J.; Krause, M.; Welch, L.

2008-01-01

Accelerator Driven Transmutation of High-Level Waste (ATW) is one possible solution to the fuel reprocessing back-end problem for the disposal of high level waste such as minor actinides (Am, Np or Cm) and long-lived fission products. International programs continue to support research towards the eventual construction and operation of a proton accelerator driven spallation neutron source coupled to a subcritical 'neutron amplifier' for more efficient HLW transmutation. This project was performed under DOE AFCI Reactor-Accelerator Coupling Experiments (RACE). A 20 MeV Electron Linac was installed in the BP no 5 cave placing neutron source adjacent to an offset reactor core to maximize neutron coupling with available systems. Asymmetric neutron injection 'wasted' neutrons due to high leakage but sufficient neutrons were available to raise reactor power to ∼100 watts. The Linac provided approximately 100 mA but only 50% reached target. The Linac cooling system could not prevent overheating at frequencies over 200 Hz. The Linac electron beam had harmonics of primary frequency and periodic low frequency pulse intensity changes. Neutron detection using fission chambers in current mode eliminated saturation dead time and produced better sensitivity. The Operation of 'dual shielded' fission chambers reduced electron noise from linac. Benchmark criticality calculation using start-up data showed that the MCNPX model overestimates reactivity. TRIGA core was loaded to just slightly supercritical by adding graphite elements and measuring reactivity of $0.037. MCNPX modeled TRIGA core with and without graphite to arrive at 'true' measured subcritical multiplication of 0.998733± 0.00069. Thus, Alpha for the UT-RACE TRIGA core was approximately 155.99 s -1 . The Stochastic Feynman-Alpha Method (SFM) accuracy was evaluated during transients and reactivity changes. SFM was shown to be a potential real-time method of reactivity determination in future ADSS but requires stable

Safety considerations of PWR's

International Nuclear Information System (INIS)

Arnold, W.H. Jr.

1977-01-01

as ECCS, as previously noted, ATWS, fire protection, turbine missile and others. Resolution of these issues has emphasized the need for utilizing probability-reliability concepts. Quantitative risk assessment is beginning to play an important role, and its importance is expected to increase, in the industry decision making process as well as the regulatory process. These tools provide a method for further quantifying the safety margins that exist in the present designs, the basis for rational decisions relative to any proposed changes in these designs as well as for any new designs aimed at improving plant safety and availability

CFD predictions of standby liquid control system mixing in lower plenum of a BWR

International Nuclear Information System (INIS)

Boyd, Christopher; Skarda, Raymond

2014-01-01

Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

Oceanographic and surface meteorological data collected from station ATW20 by University of Wisconsin-Milwaukee and assembled by Great Lakes Observing System (GLOS) in the Great Lakes region from 2014-07-01 to 2017-08-31 (NODC Accession 0123639)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — NCEI Accession 0123639 contains oceanographic and surface meteorological data in netCDF formatted files, which follow the Climate and Forecast metadata convention...

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2009-06-15

SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

International Nuclear Information System (INIS)

2009-06-01

SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

Nuclear reactor core modelling in multifunctional simulators

International Nuclear Information System (INIS)

Puska, E.K.

1999-01-01

studied to assess the possibilities for using three-dimensional cores in training simulators. The core model results have been compared with the Loviisa WWER-type plant measurement data in steady state and in some transients. Hypothetical control rod withdrawal, ejection and boron dilution transients have been calculated with various three-dimensional core models for the Loviisa WWER-440 core. Several ATWS analyses for the WWER-1000/91 plant have been performed using the three-dimensional core model. In this context, the results of APROS have been compared in detail with the results of the HEXTRAN code. The three-dimensional Olkiluoto BWR-type core model has been used for transient calculation and for severe accident re-criticality studies. The one-dimensional core model is at present used in several plant analyser and training simulator applications and it has been used extensively for safety analyses in the Loviisa WWER-440 plant modernisation project. (orig.)

Advancing Ruggedness of Nuclear Stations By Expanding Defence In Depth in Critical Areas

International Nuclear Information System (INIS)

Koshy, Thomas

2015-01-01

The nuclear industry continues to rise above the challenges it has faced over the years from external events and internal events. Fukushima event has shed light on a few vulnerabilities that could be overcome by utilizing the current state of technology. Common cause from sea water ingression was not conceived to have the entire electrical power system including AC and DC disabled beyond reasonable recovery. Rather than focusing on the solutions for lessons from Fukushima, it is better to address 'Fukushima type' events and advance the resilience of the NPPs. The effort needs to be on exploring different approaches to overcome such vulnerabilities so that a variety of solutions are available to make appropriate choices on improving NPP ruggedness based on anticipated challenges in the regions. In a technology neutral approach for light water reactors (LWR) there are 4 critical areas that are significant for ensuring nuclear safety. (1) Reactor trip, (2) Depressurization, (3) Emergency Core Cooling, and (4) Containment integrity. The reactor trip had not suffered any significant setbacks in the immediate past but provisions to address Anticipated Transients without Scram (ATWS) were generally included in most designs. While the technology has advanced, software driven/assisted trips are becoming popular and desirable. However, a diverse approach with least probability of potential interference needs to be provided in the control room and remote shutdown area to advance the ruggedness of rector trip. Depressurization is essential for passive as well as active cooling systems and therefore the approaches to de-pressurize should have more than one approach to ensure its success. In the absence of diverse approaches to de-pressurize, it is more important to consider RCS cooling capability during accidents or transients while the reactor is at a higher pressure. In the area of Emergency Core Cooling, the events history demonstrates greater success on diversity

Nuclear reactor core modelling in multifunctional simulators

Energy Technology Data Exchange (ETDEWEB)

Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

1999-06-01

studied to assess the possibilities for using three-dimensional cores in training simulators. The core model results have been compared with the Loviisa WWER-type plant measurement data in steady state and in some transients. Hypothetical control rod withdrawal, ejection and boron dilution transients have been calculated with various three-dimensional core models for the Loviisa WWER-440 core. Several ATWS analyses for the WWER-1000/91 plant have been performed using the three-dimensional core model. In this context, the results of APROS have been compared in detail with the results of the HEXTRAN code. The three-dimensional Olkiluoto BWR-type core model has been used for transient calculation and for severe accident re-criticality studies. The one-dimensional core model is at present used in several plant analyser and training simulator applications and it has been used extensively for safety analyses in the Loviisa WWER-440 plant modernisation project. (orig.) 75 refs. The thesis includes also eight previous publications by author

Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

2000-07-01

The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the

Thermal Power Of The TS-300B Refrigerator in the Aspects of Statistical Research / Moc Cieplna Chłodziarki TS-300B W Aspekcie Badań Statystycznych

Science.gov (United States)

Nowak, Bernard; Łuczak, Rafał

2015-09-01

refrigerator, both on the test stand in the manufacturer's laboratory and in the workings of underground mines. The evaluation of the measurement data distributions, as well as an analysis of the basic descriptive statistics of the mentioned variables were carried out, determining their measures of central tendency, location, dispersion and asymmetry. Artykuł dotyczy poprawy cieplnych warunków pracy w wyrobiskach górniczych kopalń podziemnych stosujących lokalne systemy chłodnicze. Rozważa się w nim skuteczność schładzania powietrza chłodziarką sprężarkową bezpośredniego działania typu TS-300B. Bardzo często, w wyniku niedotrzymania wymaganych warunków pracy wymienionego systemu chłodzenia powietrza, występują rozbieżności między prognozowanymi, a więc oczekiwanymi efektami jego pracy a rzeczywistością. Dlatego, dla poprawy skuteczności pracy tego systemu, opracowano, pod kątem efektywnego wykorzystania mocy chłodniczej parownika takiej chłodziarki, łatwe w zastosowaniu praktycznym kryteria jakości. Otrzymano je w postaci modeli statystycznych określających wpływ zmiennych niezależnych, tj. parametrów powietrza wlotowego do parownika (temperatury, wilgotności i wydatku objętościowego) oraz parametrów wody chłodzącej skraplacz (temperatury i wydatku objętościowego) na moc cieplną chłodnicy powietrza traktowaną jako zmienna zależna. Równania statystyczne opisujące pracę rozważanego systemu chłodzenia powietrza wyznaczono na podstawie wielorakiej regresji liniowej i nieliniowej. Utworzone funkcje zmodyfikowano poprzez zmianę wartości współczynników w przypadku regresji liniowej oraz współczynników i wykładników w przypadku regresji nieliniowej, przy zmiennych niezależnych. Otrzymano w ten sposób funkcje dogodniejsze w praktycznych wykorzystaniach. Korzystając z metod statystyki klasycznej oceniono jakość dopasowania funkcji regresji do danych eksperymentalnych. Porównano także wartości mocy cieplnych

Capability of APROS in design and validation of I and C systems

International Nuclear Information System (INIS)

Kantee, H.; Porkholm, K.

2004-01-01

-440 type reactor analyses since besides the control rod movement during normal operation there are four different reactor protection modes. During two protection modes control rod groups are moving individually with the velocity based on protection mode. APROS 5.02 computer code has been applied in the analyses of three RCPs trip in Loviisa 1 NPP. The main assumption was that the plant operates as designed excluding the defined failures. APROS input model of Loviisa was a best-estimate model and did not include any conservative features. These analyses indicated that from the modeling point of view the temperature measurement time constant was found to have an effect on the results. Therefore, in APROS model of Loviisa time constant of every temperature measurement is modeled. When using APROS the modeling is very easy since there is measurement component available in the code and the user only has to give the time constant as an input value. Four cases dealing with stop of three RCPs were analyzed. In the base case the reactor operated as designed. In the first variation ROM was assumed not to be available and on the second variation in addition to ROM also actuation of fast shutdown was assumed to fail. The last variation was a kind of an ATWS case where no control rod movement was assumed. Based on the results it can be concluded that if ROM fails reactivity feedback reduces thermal power until slow shutdown is actuated from high upper plenum temperature. Slow shutdown is not, however, capable to reduce upper plenum temperature fast enough below set point value and also fast shutdown is actuated 20 s later. This results in situation where fission power 'dies' and reactor power generation from this point of time onwards is based on decay heat generation. When assuming that no control rods movement takes place, reactor power decreases down to 62 % level as a result of reactivity feedback. The level is higher than desired 50% and therefore e.g. turbine bypass valves