Pin failure modeling of the A series CABRI tests

International Nuclear Information System (INIS)

Young, M.F.; Portugal, J.L.

1978-01-01

The EXPAND pin fialure model, a research tool designed to model pin failure under prompt burst conditions, has been used to predict failure conditions for several of the A series CABRI tests as part of the United States participation in the CABRI Joint Project. The Project is an international program involving France, Germany, England, Japan, and the United States and has the goal of obtaining experimental data relating to the safety of LMFBR's. The A series, designed to simulate high ramp rate TOP conditions, initially utilizes single, fresh UO 2 pins of the PHENIX type in a flowing sodium loop. The pins are preheated at constant power in the CABRI reactor to establish steady state conditions (480 w/cm at the axial peak) and then subjected to a power pulse of 14 ms to 24 ms duration

Feasibility study to restart the research reactor RA with a converted core

International Nuclear Information System (INIS)

Matausek, M.V.; Plecas, I.; Marinkovic, N.

1999-01-01

Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities, are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

Japan: The institute for the economy of energy recommends a quick re-start of nuclear reactors

International Nuclear Information System (INIS)

Anon.

2014-01-01

The Japanese Institute for the Economy of the Energy (IEEJ) considers that the sooner the nuclear reactors will re-start, the better the Japanese economy and environment will be. The 48 Japanese reactors were stopped after the Fukushima accident and their restart is linked to the implementation of new measures for reinforcing safety. Until now only 2 reactors Sendai 1 and Sendai 2 have been allowed to re-start. The procedure for the safety assessment of the reactors is slower than expected. A study shows that only 7 reactors may be allowed to re-start before march 2015 and a total of 19 units may be operating in march 2016. In this scenario 2% of the electricity will come from nuclear energy in 2014 and 15% in 2015, natural gas imports will still be necessary for the production of electricity and their global cost is estimated to reach 56 billions euros while Japan's rate of energy independence will drop by 4.6%. (A.C.)

Japan: Sendai, first reactor to restart. Sendai restart: how does it work? Japan: restart will be 'progressive'. 2015: which role for nuclear energy in Japan?

International Nuclear Information System (INIS)

Le Ngoc, Boris; Jouette, Isabelle

2015-01-01

A set of articles addresses the restart of nuclear plants in Japan. The first one presents the Sendai nuclear plant, evokes the commitment of the Japanese nuclear safety authority (the NRA) at each step of the restart process, the agreement of local populations, the loading of the nuclear fuel, a successful crisis exercise, and the benefits expected from this restart. The second article addresses the restart process with its administrative aspects, the implication of local authorities, its technical aspects, and investments made to improve nuclear safety. The third article proposes an interview of the nuclear expert of the French embassy in Tokyo. He outlines that the restart of nuclear plants will be progressive, comments how Sendai restart has been commented in the Japanese press, evokes how this restart is part of the Japanese Prime Minister's policy, evokes the role and challenges of nuclear energy in Japan for the years to come, and the role France may play. The last article discusses the role of nuclear energy in Japan in 2015: importance of the old 3E policy (Energy, Environment, Economy) which is put into question again by the Fukushima accident, creation of a new nuclear safety authority as a first step before restarting nuclear reactors

IAEA issues recommendations regarding temporary restart of Dutch reactor

International Nuclear Information System (INIS)

2009-01-01

Full text: An IAEA-led international team of nuclear reactor safety experts completed a safety review mission on 18 February at the High Flux Reactor (HFR) at Petten, in the Netherlands. The mission was conducted at the request of the Government of the Netherlands to review a set of previous evaluations made by the Dutch regulatory authority regarding the reactor's safety. The IAEA mission made a series of recommendations to enhance the safety of the year-long temporary restart. The recommendations included: - Performance of the monitoring system for leaks should be rigorously checked during the interim year of operation; - Temporary operation of the HFR cannot be extended beyond 1 March 2010; and - In case of any detected leakage from the coolant pipes, the reactor should be shut down immediately and repaired before restarting. The international team was composed of one IAEA staff member and five external experts from Argentina, Canada, France, India and South Africa. The IAEA's main conclusions and recommendations were presented in The Hague to the Ministry of Housing, Spatial Planning and the Environment and several other ministries. The team also provided a summary of its findings to the Netherlands Regulatory Authority. The team's final report will be submitted within two weeks. The HFR at Petten is one of five research reactors in the world that produces radioactive medical isotopes, used an estimated 40 million times annually for cancer treatment and the diagnosis of heart attacks. Prolonged outages at any of these five reactors have a far-reaching impact on medical treatments and diagnoses for patients around the globe. Since August 2008, the HFR reactor has been in shut-down status due to corrosion of pipes in its primary cooling circuit. The Nuclear Research and Consultancy Group (NRG), the operating organization for Petten, proposed a one-year restart of the HFR reactor, which was approved by the Dutch regulatory body. The reactor then resumed operation

Qualification and characterization of electronics of the fast neutron Hodoscope detectors using neutrons from CABRI core

Science.gov (United States)

Mirotta, S.; Guillot, J.; Chevalier, V.; Biard, B.

2018-01-01

The study of Reactivity Initiated Accidents (RIA) is important to determine up to which limits nuclear fuels can withstand such accidents without clad failure. The CABRI International Program (CIP), conducted by IRSN under an OECD/NEA agreement, has been launched to perform representative RIA Integral Effect Tests (IET) on real irradiated fuel rods in prototypical Pressurized Water Reactors (PWR) conditions. For this purpose, the CABRI experimental pulse reactor, operated by CEA in Cadarache, France, has been strongly renovated, and equipped with a pressurized water loop. The behavior of the test rod, located in that loop in the center of the driver core, is followed in real time during the power transients thanks to the hodoscope, a unique online fuel motion monitoring system, and one of the major distinctive features of CABRI. The hodoscope measures the fast neutrons emitted by the tested rod during the power pulse with a complete set of 153 Fission Chambers and 153 Proton Recoil Counters. During the CABRI facility renovation, the electronic chain of these detectors has been upgraded. In this paper, the performance of the new system is presented describing gain calibration methodology in order to get maximal Signal/Noise ratio for amplification modules, threshold tuning methodology for the discrimination modules (old and new ones), and linear detectors response limit versus different reactor powers for the whole electronic chain.

Analysis of reactivity worth for xenon poisoning during restart-up of reactor in iodine pit

International Nuclear Information System (INIS)

Li Xaofeng; Chen Wenzhen; Zhu Qian; Xu Guojun

2009-01-01

The reactivity worth of xenon poisoning and the densities of 135 I and 135 Xe were derived when the reactor was restarted up in iodine pit. Through the expressions obtained we can find the physics characteristics of reactor restarted up in iodine pit comprehensively and essentially. The results were analyzed and discussed. The reactor power before shutdown, the start-up power, the position where the reactor starts up in iodine pit, and so on, all have effect on the reactivity worth of xenon poisoning, and the different conditions can lead to totally different physics characteristics. In addition, the time when the reactor starts up in iodine pit is a very important factor for nuclear reactors safety. The conclusions are very important to the maneuverability and operation safety of ship nuclear reactors. (authors)

Safety actuator of the Cabri reactor as a function of its power and cooling fluid flow rate

International Nuclear Information System (INIS)

Bertrand, Jean; Da Costa Vieira, David; Tattegrain, Alain

1969-04-01

This report present a device which is to provide a stop command to the Cabri reactor when the rate of its power to the cooling fluid rate reaches a value determined with respect to water temperature in the circuit. The stop command is delivered by an actuator which opens a relay contact when the power reaches a specific value. The authors present the device, its characteristics, and principle. They also present the different amplifier circuits, the input and output circuits (flow rate input, temperature input, and output circuit), the energy supply, and the various adjustments

The application of modern safety criteria to restarting and operating the USDOE K-Reactor

International Nuclear Information System (INIS)

Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

1993-01-01

The United States Department of Energy's (USDOE's) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992

Safety Evaluation Report: Restart of K-Reactor, Savannah River Site

International Nuclear Information System (INIS)

1991-08-01

In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, ''Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues which need to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied a the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. This document supplements the April 1991 SER. The SER Supplement documents the resolution of several of the open items identified in the SER. Only those issues (sections) for which at least one open item identified in the SER has now been closed are addressed in this Supplement. Additionally, some SER sections had no open items identified. Therefore, this Supplement does not include all sections that were addressed in the SER. If there are any open items remaining to be resolved for the sections included in this Supplement, that is so identified at the end of the section. The resolution of those remaining open times, and all remaining open items for those SER sections not included in this first Supplement, will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed

Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

International Nuclear Information System (INIS)

Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

1990-01-01

The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

Study of the linearity of CABRI experimental ionization chambers during RIA transients

Science.gov (United States)

Lecerf, J.; Garnier, Y.; Hudelot, JP.; Duc, B.; Pantera, L.

2018-01-01

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center and funded by the French Nuclear Safety and Radioprotection Institute (IRSN). For the purpose of the CABRI International Program (CIP), operated and managed by IRSN under an OECD/NEA framework it has been refurbished since 2003 to be able to provide experiments in prototypical PWR conditions (155 bar, 300 °C) in order to study the fuel behavior under Reactivity Initiated Accident (RIA) conditions. This paper first reminds the objectives of the power commissioning tests performed on the CABRI facility. The design and location of the neutron detectors monitoring the core power are also presented. Then it focuses on the different methodologies used to calibrate the detectors and check the consistency and co-linearity of the measurements. Finally, it presents the methods used to check the linearity of the neutron detectors up to the high power levels ( 20 GW) reached during power transients. Some results obtained during the power tests campaign are also presented.

The CABRI facility: Implementation of a pressurized water loop and related safety review

International Nuclear Information System (INIS)

Cabrillat, J.C.; Maegey, M.; Bourguignon, D.; Miachon, G.; Forestier, F.; Coulon, J.P.; Faury, M.

2003-01-01

The CABRI reactor operated by CEA at the CADARACHE Nuclear Center in France provided the condition for safety studies on nuclear fuel. Initially designed to support investigations on Fast Reactor fuel, large modifications are underway to provide representative conditions for studies on Light Water Fuel types. A general overview of these modifications, the related safety review and supporting studies are described in the paper. (author)

The behaviour of irradiated fuel under RIA transients: Interpretation of the CABRI experiments

International Nuclear Information System (INIS)

Papin, J.; Rigat, H.; Breton, J.P.; Schmitz, F.

1996-01-01

Paper presents the results of investigation of highly irradiated PWR fuel behaviour under fast power transients conducted in a sodium loop of CABRI reactor, as well as the results on development and validation of computer code SCANAIR. (author). 8 refs, 9 figs, 2 tabs

Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

2009-01-01

In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

Review of Savannah River Site K Reactor inservice inspection and testing restart program

International Nuclear Information System (INIS)

Anderson, M.T.; Hartley, R.S.; Kido, C.

1992-09-01

Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

Analysis of ASTEC-Na capabilities for simulating a loss of flow CABRI experiment

International Nuclear Information System (INIS)

Flores y Flores, A.; Matuzas, V.; Perez-Martin, S.; Bandini, G.; Ederli, S.; Ammirabile, L.; Pfrang, W.

2016-01-01

Highlights: • ASTEC-Na results for CABRI BI1 test have been compared with experimental data. • The ASTEC-Na calculations reached the boiling onset within the error bar of the test. • The coolant axial profile in ASTEC-Na fit almost perfectly with the experimental data. • All the calculations have a good agreement with the two-phase front downwards. • All the calculations have a worse agreement with the two-phase front upwards. - Abstract: This paper presents simulation results of the CABRI BI1 test using the code ASTEC-Na, currently under development, as well as a comparison of the results with available experimental data. The EU-JASMIN project (7th FP of EURATOM) centres on the development and validation of the new severe accident analysis code ASTEC-Na (Accident Source Term Evaluation Code) for sodium-cooled fast reactors whose owner and developer is IRSN. A series of experiments performed in the past (CABRI/SCARABEE experiments) and new experiments to be conducted in the new experimental sodium facility KASOLA have been chosen to validate the developed ASTEC-Na code. One of the in-pile experiments considered for the validation of ASTEC-Na thermal–hydraulic models is the CABRI BI1 test, a pure loss-of-flow transient using a low burnup MOX fuel pin. The experiment resulted in a channel voiding as a result of the flow coast-down leading to clad melting. Only some fuel melting took place. Results from the analysis of this test using SIMMER and SAS-SFR codes are also presented in this work to check their suitability for further code benchmarking purposes.

The 'SURA' fast reactor program

International Nuclear Information System (INIS)

Anon.

1979-01-01

The Commissariat a l'Energie Atomique's SURA program on fast reactor safety consists of two specific testing programs on fastbreeder reactor safety: the Cabri and Scarabee programs. Both Cabri and Scarabee are examples of multinational research collaboration. The CEA and the Karlsruhe Nuclear Research Center are each covering half of the construction costs. Britain, the US and Japan are also due to participate in these experiments. The aim of the programs is to examine the behaviour of fuel in sodium cooled fast reactors. The Cabri program consists of setting off a reactivity accident in a power reactor core which is cooled with liquid sodium, such an accident occurring after a sharp increase in reactivity or as a result of the pump suddenly breaking down without there at the same time being any fall in the control rods. In 1967 the Commissariat a l'Energie Atomique started its Scarabee research program which is trying to analyse the sort of things that can go wrong with fuel cooling systems and what the consequences can be [fr

Results of CABRI start-up measurements with the Zircaloy loop

International Nuclear Information System (INIS)

Kussmaul, G.; Rongier, P.

1983-02-01

After the change of the test loop in the CABRI core a start-up programme has been performed to determine the gain of coupling and the He-3 pressure-dependent driver core energy release for transients from different initial power levels and modified injection rates. This report gives the results of thirty tests and the means to adjust the experimental conditions of future CABRI tests

Evaluation of the Steel Creek ecosystem in relation to the proposed restart of L reactor

International Nuclear Information System (INIS)

Smith, M.H.; Sharitz, R.R.; Gladden, J.B.

1981-10-01

Information is presented on the following subjects: habitat and vegetation, the avifauna, semi-aquatic and terrestrial vertebrates, and aquatic communities of Steel Creek, species of special concern, and radiocesium in Steel Creek. Two main goals of the study were the compilation of a current inventory of the flora and fauna of the Steel Creek ecosystem and an assessment of the probable impacts of radionuclides, primarily 137 Cs, that were released into Steel Creek during earlier reactor operations. Although a thorough evaluation of the impacts of the L reactor restart is impossible at this time, it is concluded that the effects on the Steel Creek ecosystem will be substantial if no mitigative measures are taken

Keefektifan Pembelajaran Model Quantum Teaching Berbantuan Cabri 3D Terhadap Kemampuan Pemecahan Masalah

Directory of Open Access Journals (Sweden)

E. Widiyaningsih

2013-06-01

Full Text Available Penelitian ini bertujuan untuk mengetahui keefektifan pembelajaran model Quantum Teaching berbantuan Cabri 3D terhadap kemampuan pemecahan masalah peserta didik. Populasi dalam penelitian ini adalah peserta didik kelas X SMA Negeri 11 Semarang tahun pelajaran 2012/ 2013. Ada dua kelas sampel yaitu kelas eksperimen yang diterapkan pem-belajaran matematika menggunakan model pembelajaran Quantum Teaching berbantuan Cabri 3D dan kelas kontrol yang diterapkan pembelajaran Direct Interaction(DI. Hasil menunjukkan bahwa hasil belajar peserta didik pada aspek kemampuan pemecahan masalah menggunakan model pembelajaran Quantum Teaching mencapai KKM individu dan klasikal, rata-rata hasil belajar peserta didik pada aspek kemampuan pemecahan masalah pada kelas eksperimen lebih baik dari kelas kontrol. Berdasarkan pembahasan di atas dapat disimpulkan bahwa pembelajaran menggunakan model Quantum Teaching berbantuan Cabri 3D efektif terhadap kemampuan pemecahan masalah peserta didik.   This paper purpose of this study was to determine learning using the Quantum Teaching model exploration with Cabri 3D is effective to student’s problem solving ability. The population in this study was students of grade X Eleven Senior High School of Semarang academic year 2012/2013. There were two sample classes; those were an experimental class that applied mathematics learning using the Quantum Teaching model exploration with Cabri 3D and control class that applied Direct Instruction (DI learning. The results showed that the learning outcomes of students in problem solving ability aspects using the Quantum Teaching model exploration with Cabri 3D  can achieve individual and classical mastery learning, the percentage of learning outcomes of student’s problem solving ability in experiment class  better than  learning outcomes of student’s problem solving ability in control class, the average of learning outcomes of student’s problem solving ability in experiment

Mental health effects of the Three Mile Island nuclear reactor restart.

Science.gov (United States)

Dew, M A; Bromet, E J; Schulberg, H C; Dunn, L O; Parkinson, D K

1987-08-01

Controversy over potential mental health effects of the Three Mile Island Unit-1 restart led the authors to examine prospectively the pattern of psychiatric symptoms in a sample of Three Mile Island area mothers of young children. Symptom levels after restart were elevated over previous levels; a sizable subcohort of the sample reported relatively serious degrees of postrestart distress. History of diagnosable major depression and generalized anxiety following the Three Mile Island accident, plus symptoms and beliefs about personal risk prior to the restart, best predicted postrestart symptoms.

Mental health effects of the Three Mile Island nuclear reactor restart

International Nuclear Information System (INIS)

Dew, M.A.; Bromet, E.J.; Schulberg, H.C.; Dunn, L.O.; Parkinson, D.K.

1987-01-01

Controversy over potential mental health effects of the Three Mile Island Unit-1 restart led the authors to examine prospectively the pattern of psychiatric symptoms in a sample of Three Mile Island area mothers of young children. Symptom levels after restart were elevated over previous levels; a sizable subcohort of the sample reported relatively serious degrees of postrestart distress. History of diagnosable major depression and generalized anxiety following the Three Mile Island accident, plus symptoms and beliefs about personal risk prior to the restart, best predicted postrestart symptoms

Cabri 3D - assisted collaborative learning to enhance junior high school students’ spatial ability

Science.gov (United States)

Muntazhimah; Miatun, A.

2018-01-01

The main purpose of this quasi-experimental study was to determine the enhancement of spatial ability of junior high school students who learned through Cabri-3D assisted collaborative learning. The methodology of this study was the nonequivalent group that was conducted to students of the eighth grade in a junior high school as a population. Samples consisted one class of the experimental group who studied with Cabri-3D assisted collaborative learning and one class as a control group who got regular learning activity. The instrument used in this study was a spatial ability test. Analyzing normalized gain of students’ spatial ability based on mathemathical prior knowledge (MPK) and its interactions was tested by two-way ANOVA at a significance level of 5% then continued with using Post Hoc Scheffe test. The research results showed that there was significant difference in enhancement of the spatial ability between students who learnt with Cabri 3D assisted collaborative learning and students who got regular learning, there was significant difference in enhancement of the spatial ability between students who learnt with cabri 3D assisted collaborative learning and students who got regular learning in terms of MPK and there is no significant interaction between learning (Cabri-3D assisted collaborative learning and regular learning) with students’ MPK (high, medium, and low) toward the enhancement of students’ spatial abilities. From the above findings, it can be seen that cabri-3D assisted collaborative learning could enhance spatial ability of junior high school students.

Mechanical energy release in CABRI-2 experiments with Viggen-4 fuel pins

International Nuclear Information System (INIS)

Wolff, J.

1993-07-01

The results of mechanical energy release evaluations in CABRI-2 experiments with Viggen-4 fuel pins (12 atom % burnup) are described. In general the experience gained by the CABRI-1 experiments is confirmed. Those physical phenomena are enhanced which are influenced by the release of fission products. Especially the late blow-out of pressurized fission gases from the lower test pin plenum led to large flow variations. The corresponding mechanical power releases are low

Experiences of graduate students: Using Cabri as a visualization tool in math education

Directory of Open Access Journals (Sweden)

Çiğdem Gül

2014-12-01

Full Text Available Through the use of graphic calculators and dynamic software running on computers and mobile devices, students can learn complex algebraic concepts. The purpose of this study is to investigate the experiences of graduate students using Cabri as a visualization tool in math education. The qualitative case study was used in this study. Five students from graduate students studying at the non-thesis math program of a university located in the Blacksea region were the participant of the study. As a dynamic learning tool, Cabri provided participants an environment where participants visually discovered the geometry. It was concluded that dynamic learning tools like Cabri has a huge potential for teaching visually the challenging concepts that students struggle to image. Further research should investigate the potential plans for integrating the use of dynamic learning software into the math curriculum

Review of fast reactor activities in Italy, April 1978

International Nuclear Information System (INIS)

Pierantoni, F.

1978-01-01

In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments

Review of fast reactor activities in Italy, April 1978

Energy Technology Data Exchange (ETDEWEB)

Pierantoni, F [CNEN Fast Reactor Programme, Bologna (Italy)

1978-07-01

In summary, the Italian fast reactor programme was developing in the following directions: PEC reactor, SUPEPHENIX reactor and long-term research and development work. Research was related to sodium technology, steam generators development, pumps, tests on mechanics and thermal insulation, core fluid dynamics, noise analysis, studies of oxide and carbide fuels, reactor safety, CABRI and SCARABEE experiments.

Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

International Nuclear Information System (INIS)

Sato, Ikken; Lemoine, Francette; Struwe, Dankward

2004-01-01

In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

Restarting TMI unit one: social and psychological impacts

International Nuclear Information System (INIS)

Sorensen, J.; Soderstrom, J.; Bolin, R.; Copenhaver, E.; Carnes, S.

1983-12-01

A technical background is provided for preparing an environmental assessment of the social and psychological impacts of restarting the undamaged reactor at Three Mile Island (TMI). Its purpose is to define the factors that may cause impacts, to define what those impacts might be, and to make a preliminary assessment of how impacts could be mitigated. It does not attempt to predict or project the magnitude of impacts. Four major research activities were undertaken: a literature review, focus-group discussions, community profiling, and community surveys. As much as possible, impacts of the accident at Unit 2 were differentiated from the possible impacts of restarting Unit 1. It is concluded that restart will generate social conflict in the TMI vicinity which could lead to adverse effects. Furthermore, between 30 and 50 percent of the population possess characteristics which are associated with vulnerability to experiencing negative impacts. Adverse effects, however, can be reduced with a community-based mitigation strategy

Restarting TMI unit one: social and psychological impacts

Energy Technology Data Exchange (ETDEWEB)

Sorensen, J.; Soderstrom, J.; Bolin, R.; Copenhaver, E.; Carnes, S.

1983-12-01

A technical background is provided for preparing an environmental assessment of the social and psychological impacts of restarting the undamaged reactor at Three Mile Island (TMI). Its purpose is to define the factors that may cause impacts, to define what those impacts might be, and to make a preliminary assessment of how impacts could be mitigated. It does not attempt to predict or project the magnitude of impacts. Four major research activities were undertaken: a literature review, focus-group discussions, community profiling, and community surveys. As much as possible, impacts of the accident at Unit 2 were differentiated from the possible impacts of restarting Unit 1. It is concluded that restart will generate social conflict in the TMI vicinity which could lead to adverse effects. Furthermore, between 30 and 50 percent of the population possess characteristics which are associated with vulnerability to experiencing negative impacts. Adverse effects, however, can be reduced with a community-based mitigation strategy.

Evaluation of the Steel Creek ecosystem in relation to the proposed restart of the L-reactor

International Nuclear Information System (INIS)

Smith, M.H.; Sharitz, R.R.; Gladden, J.B.

1982-10-01

This report summarizes the findings of slightly more than one year's study of the Steel Creek ecosystem. Generally, the findings have allowed us to refine our understanding of the structural and functional organization of the Steel Creek ecosystem which is an essential prerequisite for predicting the impacts associated with L-reactor restart. Reanalysis of the Steel Creek plant community relationships using 1981 aerial photography revealed that this component of the delta ecosystem continues to change as a result of natural successional processes. The major detectable changes have occurred on the more elevated portions of Steel Creek delta where coverage by woody species (especially willow) is continuing to increase. This successional woody community is invading areas previously dominated by persistent herbaceous species such as cut grass. Eleven vegetation associations were identified in the Steel Creek delta area, including two associations that were not apparently affected by the earlier reactor operations

Overview on phenomena of mechanical energy release in the CABRI-experiments

International Nuclear Information System (INIS)

Wolff, J.

1989-01-01

Mechanical energy release phenomena observed in the CABRI-experiments are overviewed and discussed. Intensifying and mitigating effects are identified. They are caused by fission gases, inertia effects, pressure dissipation and fissile power

K-Reactor readiness

International Nuclear Information System (INIS)

Rice, P.D.

1991-01-01

This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan

Students’ Spatial Ability through Open-Ended Approach Aided by Cabri 3D

Science.gov (United States)

Priatna, N.

2017-09-01

The use of computer software such as Cabri 3D for learning activities is very unlimited. Students can adjust their learning speed according to their level of ability. Open-ended approach strongly supports the use of computer software in learning, because the goal of open-ended learning is to help developing creative activities and mathematical mindset of students through problem solving simultaneously. In other words, creative activities and mathematical mindset of students should be developed as much as possible in accordance with the ability of spatial ability of each student. Spatial ability is the ability of students in constructing and representing geometry models. This study aims to determine the improvement of spatial ability of junior high school students who obtained learning with open-ended approach aided by Cabri 3D. It adopted a quasi-experimental method with the non-randomized control group pretest-posttest design and the 2×3 factorial model. The instrument of the study is spatial ability test. Based on analysis of the data, it is found that the improvement of spatial ability of students who received open-ended learning aided by Cabri 3D was greater than students who received expository learning, both as a whole and based on the categories of students’ initial mathematical ability.

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

Science.gov (United States)

Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno

2018-01-01

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

Directory of Open Access Journals (Sweden)

Clamens Olivier

2018-01-01

Full Text Available CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

Savannah River Site peer evaluator standards: Operator assessment for restart

International Nuclear Information System (INIS)

1990-01-01

Savannah River Site has implemented a Peer Evaluator program for the assessment of certified Central Control Room Operators, Central Control Room Supervisors and Shift Technical Engineers prior to restart. This program is modeled after the nuclear Regulatory Commission's (NRC's) Examiner Standard, ES-601, for the requalification of licensed operators in the commercial utility industry. It has been tailored to reflect the unique differences between Savannah River production reactors and commercial power reactors

The restart of Belgium reactors of Doel 3 and Tihange 2. Doel 3 and Tihange 2: indications of defects in vessel steel

International Nuclear Information System (INIS)

2015-01-01

In a first part, an IRSN report comments the issue of restarting some Belgium reactors after the detection of defects (due to the presence of hydrogen) in the vessel steel of reactors during the third decennial inspection by Electrabel. The report describes the procedure followed by Electrabel and the Belgium nuclear authority (AFCN) to confirm that the detected defects were not harmful. It comments the defect detection and characterisation, the origin and potential evolution of defects, the assessment of mechanical characteristics of some components, the assessment of the defect harmfulness in terms of failure risk, and additional measurements. The second part contains the AFCN report which addresses: the chronology and scientific context, the actors, the situation of other Belgium reactors, an indication of published reports and press releases

Risk perception in an interest group context: an examination of the TMI restart issue

International Nuclear Information System (INIS)

Soderstrom, E.J.; Sorensen, J.H.; Copenhaver, E.D.; Carnes, S.A.

1984-01-01

Human response to environmental hazards and risks has been the subject of considerable research by social scientists. Work has traditionally focused on either individual response to the risks of an ongoing or future threat (hazards research), or group and organizational response to a specific disaster event (disaster research). As part of a larger investigation of the restart of the Unit 1 reactor at Three Mile Island (TMI), the response of interest groups active in the restart issue to the continued threat of TMI and to future risks due to restart was examined. After reviewing the restart issue in general, the local dimensions of the restart issue from interest group perspectives are discussed. A method for defining appropriate issues at the community level is reviewed. Differences in the perceived local impacts of alternative decisions, and systems of beliefs associated with differing perceptions are discussed. Finally, the implications of interest group versus individual perceptions of local issues for decision making about TMI, in particular, and about technological hazards management, in general, are discussed. Associated implications for determining socially acceptable risk levels are identified

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

2012 review of French research reactors

International Nuclear Information System (INIS)

Estrade, Jerome

2013-01-01

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

Effect of PWR Re-start ramp rate on pellet-cladding interactions

International Nuclear Information System (INIS)

Yagnik, S.K.; Chang, B.C.; Sunderland, D.J.

2005-01-01

To mitigate pellet-cladding interaction (PCI) leading to fuel rod failures, fuel suppliers specify reactor power ramp rate limitations during reactor start-up after an outage. Typical re-start ramp rates are restricted and range between 3-4% per hour of full reactor power above a threshold power level. Relaxation of threshold power and ramp rate restrictions has the potential to improve plant economics. The paper will compare known re-start power ascension procedures employed in the US, German, French and Korean PWRs after a refuelling outage. A technical basis for optimising power ascension procedures during reactor start-up can be developed using analytical modelling. The main objective of the modelling is to determine the potential for PCI failure for various combinations of threshold power levels and ramp rate levels. A key element of our analysis is to estimate the decrease in margin to cladding failure by ISCC based on a time-temperature-stress failure criterion fashioned Act a cumulative cladding damage index. The analysis approach and the cladding damage model will be described and the results from three case studies based on the FALCON fuel rod behaviour code will be reported. We conclude that the PCI behaviour is more affected by ramp rate and threshold power than by the fuel design and that the fuel power history is the most important parameter. (authors)

Situaciones a-didácticas para la enseñanza de la homotecia utilizando CabriLM como medio

OpenAIRE

Pérez, Luis

2013-01-01

Se presenta una serie de situaciones a-didácticas, en las que los alumnos interactúan con el software CabriLM como medio, y gracias a dichas interacciones construyen conocimientos relacionados con el concepto de homotecia. El funcionamiento geométrico de los objetos de CabriLM garantiza que los fenómenos visuales corresponden a propiedades teóricas, y las posibilidades que ofrece para controlar las interacciones permiten introducir restricciones para bloquear estrategias no matemáticas para r...

Restarts in Conversation and Literature.

Science.gov (United States)

Person, Raymond F., Jr.

1996-01-01

Analyzes restarts, a common feature of conversation, in literary discourse. The term "restart" refers to the repetition of a word or words within an utterance by the same speaker. Restarts in literary discourse are of two types: (1) those produced by the characters in their "real" narrative world and (2) those produced by the narrators themselves.…

Reactivity feedback evaluation of material relocations in the CABRI-1 experiments with fuel worth distributions from SNR-300

International Nuclear Information System (INIS)

Royl, P.; Pfrang, W.; Struwe, D.

1991-01-01

The fuel relocations from the CABRI-1 experiments with irradiated fuel that had been evaluated from the hodoscope measurements were used together with fuel reactivity worth distributions from the SNR-300 to estimate the reactivity effect which these motions would have if they occurred in SNR-300 at the same relative distance to the peak power as in CABRI. The procedure for the reactivity evaluation is outlined including the assumptions made for fuel mass conservation. The results show that the initial fuel motion yields always negative reactivities. They also document the mechanism for a temporary reactivity increase by in-pin fuel flow in some transient overpower tests. This mechanism, however, never dominates, because material accumulates always sufficiently above the peak power point. Thus, the late autocatalytic amplifications of voiding induced power excursions by compactive in-pin fuel flow, that had been simulated in bounding loss of flow analyses for SNR-300, have no basis at all when considering the results from the CABRI-1 experiments

Programme of CABRI start-up measurements with the Zircaloy loop

International Nuclear Information System (INIS)

Kussmaul, G.; Rongier, P.

1981-06-01

After installation and operational tests of the CABRI Zircaloy loop, a start-up test programme will be carried out to determine the new coupling value between the driver core and the test pin and the reactivity dependent driver core energy release for transients from different power levels and modified injection rates. The purpose of the tests and the test programme itself are described in the report

Reproducible Data Processing Research for the CABRI R.I.A. experiments Acoustic Emission signal analysis

Energy Technology Data Exchange (ETDEWEB)

Pantera, Laurent [CEA, DEN, CAD/DER/SRES/LPRE, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Issiaka Traore, Oumar [Laboratory of Machanics and Acoustics (LMA) CNRS, 13402 Marseille (France)

2015-07-01

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission (CEA) designed to study the behaviour of fuel rods at high burnup under Reactivity Initiated Accident (R.I.A.) conditions such as the scenario of a control rod ejection. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream from the test device. Studies carried out on the last fourteen tests showed the interest in carrying out temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena. We want presently to return to this rich data in order to have a new point of view by applying modern signal processing methods. Such an antecedent works resumption leads to some difficulties. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not clear in reproducing from the former studies since the people who were in charge of the original work have left the laboratory and it is not easy when time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup. Thus we decided to consolidate the availability of the data and its manipulation in order to provide a robust data processing workflow to the experimentalists before doing any further investigations. To tackle this issue of strong links between data, treatments and the generation of documents, we adopted a Reproducible Research paradigm. We shall first present the tools chosen in our laboratory to implement this workflow and, then we shall describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the last fourteen CABRI R.I.A. tests. (authors)

The implicit restarted Arnoldi method, an efficient alternative to solve the neutron diffusion equation

Energy Technology Data Exchange (ETDEWEB)

Verdu, G.; Miro, R. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universidad Politecnica de Valencia, Valencia (Spain); Vidal, V. [Departamento de Sistemas Informaticos y Computacion, Universidad Politecnica de Valencia, Valencia (Spain)

1999-05-01

To calculate the neutronic steady state of a nuclear power reactor core and its subcritical modes, it is necessary to solve a partial eigenvalue problem. In this paper, an implicit restarted Arnoldi method is presented as an advantageous alternative to classical methods as the Power Iteration method and the Subspace Iteration method. The efficiency of these methods, has been compared calculating the dominant Lambda modes of several configurations of the Three Mile Island reactor core.

Shutdown channels and fitted interlocks in atomic reactors

International Nuclear Information System (INIS)

Furet, J.; Landauer, C.

1968-01-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

2000-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1998-01-01

Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1996-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

Anticipated transport of Cs-137 from Steel Creek following L-Area restart

International Nuclear Information System (INIS)

Hayes, D.W.

1982-01-01

Heat exchanger cooling water, spent fuel storage basin effluents, and process water from P and L-Reactor Areas were discharged to Steel Creek beginning in 1954. Cs-137 was the most significant radionuclide discharged to the environs. Once the Cs-137 was discharged from P and L-Area reactors to Steel Creek, it became associated with silt and clay in the Steel Creek system. After its association with the silt and clay, the Cs-137 becomes part of the sediment transport process and undergoes continual deposition-resuspension in the stream system. This report discusses the expected fate and transport of Cs-137 currently present in the Steel Creek system after L-Reactor restart

First analysis of AGS0, LT2 and E9 CABRI tests with the new SFR safety code ASTEC-Na

International Nuclear Information System (INIS)

Perez-Martin, Sara; Bandini, Giacomino; Matuzas, Vaidas; Buck, Michael; Girault, Nathalie

2015-01-01

Within the framework of the European JASMIN project, the ASTEC-Na code is being developed for safety analysis of severe accidents in SFR. In the first phase of validation of the ASTEC-Na fuel thermo-mechanical models three in-pile tests conducted in the CABRI experimental reactor have been selected to be analysed. We present here the preliminary results of the simulation of two Transient Over Power tests and one power ramp test (AGS0, LT2 and E9, respectively) where no pin failure occurred during the transient. We present the comparison of ASTEC-Na results against experimental data and other safety code results for the initial steady state conditions prior to the transient onset as well as for the fuel pin behaviour during the transients. (author)

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-06-01

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO₂) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

Some particular aspects of control in nuclear power reactors

International Nuclear Information System (INIS)

Vathaire, F. de; Vernier, Ph.; Pascouet, A.

1964-01-01

This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

Colombo cabri or vegetarian meal: wherein lies the power?

Directory of Open Access Journals (Sweden)

Meritxell Martín-i-Pardo

2006-05-01

Full Text Available “Colombo Cabri or Vegetarian Meal” montre que certains aliments sont utilisés pour définir deux sectes Hindouistes concurrentes en Guadeloupe, dans les Antilles françaises. Ces groupes expriment des identités différentes concurrentes en termes de visibilité et de revendication de pouvoir pour contrôler l’avenir de l’Hindouisme sur l’île. Ces deux entités, identifiées à juste titre comme Hindouistes « traditionalistes » d’une part et « globalistes » d’autre part, mettent en œuvre une rhétorique pour légitimer leurs différentes revendications qui passe par l’adoption ou le rejet du « colombo », un curry de viandes marinées dans cette sauce, comme repas rituel de la secte dont le discours vise à dicter le droit chemin de la communauté Hindouiste de l’île.“Colombo Cabri or Vegetarian Meal” argues that certain foods are used to configure two competing sectarian Hindu groups in Guadeloupe, French West Indies. These groups configure different identities to compete for the attention and claim effective power to control the future of Hinduism on the island. What are appropriately identified as “traditionalist” and “globalist” Hindus define a rhetoric for legitimating their different claims by appropriating or rejecting “colombo,” a curry of meats simmered in this sauce, as the ritual meal for the sect whose narrative rightly claims to define the correct path for the Hindu community on the island.

Additional information for impact response of the restart safety rods

International Nuclear Information System (INIS)

Yau, W.W.F.

1991-01-01

WSRC-RP-91-677 studied the structural response of the safety rods under the conditions of brake failure and accidental release. It was concluded that the maximum impact loading to the safety rod is 6020 pounds based on conservative considerations that energy dissipation attributable to fluid resistance and reactor superstructure flexibility. The staffers of the Defense Nuclear Facility Safety Board reviewed the results and inquired about the extent of conservatism. By request of the RESTART team, I reassessed the impact force due to these conservative assumptions. This memorandum reports these assessments

Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kroeger, P.G.

1992-01-01

During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

An experimental study of the behaviour of fission products following an accident on a swimming pool reactor

International Nuclear Information System (INIS)

Dadillon, J.

1976-11-01

In the estimation of nuclear risks connected with the running of a reactor an essential factor, sometimes neglected because insufficiently known, is the knowledge of the type, amount and behaviour of the contamination actually released inside the containment in the case of an accident. In the special case of swimming pool reactors the cooling fluid proves to be a very efficient barrier against contamination. Three experiments were carried out in the reactor CABRI, during which several fuel element plates were melted inside the core itself. (Author)

Restarting Automata with Auxiliary Symbols and Small Lookahead

DEFF Research Database (Denmark)

Schluter, Natalie Elaine

2012-01-01

We present a study on lookahead hierarchies for restarting automata with auxiliary symbols and small lookahead. In particular, we show that there are just two different classes of languages recognised by RRWW automata, through the restriction of lookahead size. We also show that the respective...... (left-) monotone restarting automaton models characterise the context-free languages and that the respective right-left-monotone restarting automata characterise the linear languages both with just lookahead length 2....

On the entrance distribution in RESTART simulation

NARCIS (Netherlands)

Garvels, M.J.J.; Kroese, Dirk

The RESTART method is a widely applicable simulation technique for the estimation of rare event probabilities. The method is based on the idea to restart the simulation at certain intermediate stages, in order to generate more occurrences of the rare event. In many cases we are interested in the

Mechanisms of bacterial DNA replication restart

Science.gov (United States)

Windgassen, Tricia A; Wessel, Sarah R; Bhattacharyya, Basudeb

2018-01-01

Abstract Multi-protein DNA replication complexes called replisomes perform the essential process of copying cellular genetic information prior to cell division. Under ideal conditions, replisomes dissociate only after the entire genome has been duplicated. However, DNA replication rarely occurs without interruptions that can dislodge replisomes from DNA. Such events produce incompletely replicated chromosomes that, if left unrepaired, prevent the segregation of full genomes to daughter cells. To mitigate this threat, cells have evolved ‘DNA replication restart’ pathways that have been best defined in bacteria. Replication restart requires recognition and remodeling of abandoned replication forks by DNA replication restart proteins followed by reloading of the replicative DNA helicase, which subsequently directs assembly of the remaining replisome subunits. This review summarizes our current understanding of the mechanisms underlying replication restart and the proteins that drive the process in Escherichia coli (PriA, PriB, PriC and DnaT). PMID:29202195

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

Shutdown channels and fitted interlocks in atomic reactors; Chaines de securite et verrouillages installes sur les piles atomiques

Energy Technology Data Exchange (ETDEWEB)

Furet, J; Landauer, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [French] Ce catalogue est compose d'un ensemble de tableaux (a raison de un tableau par pile) donnant les renseignements suivants: nombre et nature des detecteurs, dynamique des chaines, nature de l'electronique associee, seuils provoquant des actions de securite, verrouillages installes. Ces fiches ont ete etablies en vue de l'examen de la securite des piles par la 'Sous-Commission de Surete des Piles', et tiennent compte des decisions de celle-ci. Les reacteurs concernes sont: Azur, Cabri, Cator-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, et Ulysse. (auteurs)

Recommendations for a restart of Molten Salt Reactor development

International Nuclear Information System (INIS)

Moir, R. W.

2007-01-01

The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 2 32U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

Design study of eventual core conversion for the research reactor RA

International Nuclear Information System (INIS)

Matausek, M. V.; Marinkovic, N.

1998-01-01

Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective, multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

Recommendations for a restart of molten salt reactor development

International Nuclear Information System (INIS)

Moir, R.W.

2008-01-01

The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often-cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. Strong incentives for the molten salt reactor design are its good fuel utilization, good economics, amazing fuel flexibility and promised large benefits. It can: - use thorium or uranium; - be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have an epithermal neutron spectrum; - fission uranium isotopes and plutonium isotopes; - produces less long-lived wastes than today's reactors by a factor of 10-100; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 deg. C if carbon composites are successfully developed. Enhancing 232 U content in the uranium to over 500 ppm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR are enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high-enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/year base program for 10 years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/year over 20 years). A benefit of liquid fuel is that

Analysis of ULOF accident in Monju reflecting the knowledge from CABRI in-pile experiments and others

International Nuclear Information System (INIS)

Sato, Ikken; Tobita, Yoshiharu; Suzuki, Tohru; Kawada, Ken-ichi; Fukano, Yoshitaka; Fujita, Satoshi; Kamiyama, Kenji; Ishikawa, Makoto; Nonaka, Nobuyuki; Usami, Shin

2007-05-01

In the safety evaluation in the original licensing procedure of the prototype FBR Monju, mechanical energy release during an unprotected loss-of-flow (ULOF) event, one of technically inconceivable events postulated beyond design basis, was evaluated and shown that the radiological consequence is suitably limited. Since the sodium leakage accident in the secondary heat-transport system, the Monju plant has been kept under the stand-by condition for more than ten years and the composition of fuel in the current reactor core has changed due to decay of a plutonium isotope into americium. This change in fuel composition affects the neutronic characteristics of the core, requiring assessment of its effect on safety. In this study, event sequences of ULOF were analyzed to assess the effect of the change in neutronic characteristics on the mechanical energy release during ULOF. A significant advance has been made in the safety analysis codes reflecting the knowledge obtained through extensive safety research programs in the last decades, notably the CABRI in-pile experiments. The present results with the advanced analysis codes showed that the mechanical energy release in the current Monju core with the changed neutronic characteristics would not exceed the value evaluated formerly. (author)

Correlation between fuel rack sticking and unintentional re-starting of EDG

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Cheol; Chung, Woo geun; Kang, Seung Hee; Kim, Myeong hoon [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Emergency Diesel Generator (EDG) was being tested after overhaul maintenance. While the EDG was running at the rated speed (450 rpm), an operator pressed the manual stop button. But the EDG failed to stop and unintentionally started again. After the unintentional re-start, the EDG maintained running speed of 340 rpm. In the category of a governing system, this paper analyzes the cause of unintentional restart of the EDG that unintentionally re-started and maintained a speed at 340 rpm. The results of the analysis were then verified by a test run. Finally, we identified a correlation between fuel rack sticking and unintentional re-starting of the EDG. An analysis was conducted to confirm the cause of an EDG which was unintentionally restarting and running at 340rpm (rated speed is 450 rpm). Through a test run, it was confirmed that the results of the analysis are correct. The cause of the EDG unintentionally restarting was that it still rotated at 55 rpm over the minimum starting speed at the moment when the shutdown cylinder stopped blocking the fuel, because of a stuck fuel rack at the R7 cylinder. At the same time, the fuel that had been supplied into the cylinders (combustion chamber) by the governing system exploded and the EDG restarted unintentionally.

El uso del Cabri en la demostracion de teoremas de la geometria euclidiana

OpenAIRE

Fuentes, Fabio; Castañez, Orlando; Gordillo, Luís; Mestre, Gelis

2009-01-01

Con el presente taller se propone utilizar el CABRI PLUS II para abordar las demostraciones de algunos teoremas de la geometría euclidiana, asociados con ángulos, triángulos y cuadriláteros, temática que es abordada en la básica primaria y secundaria, al igual que en la geometría euclidiana de la licenciatura de matemáticas de las Universidades de Colombia. La utilización de las demostraciones de teoremas es considerada una potente herramienta que ayuda a mejorar los procesos de enseñanza y d...

Methods used and kind of results obtained in SCARABEE facility about fuel and clad motion diagnostics. Previsional development for SCARABEE N and CABRI facilities

International Nuclear Information System (INIS)

Bardy, J.; Manent, G.; Roche, L.; Tattegrain, A.

1975-01-01

Techniques are described for planned LMFBR loss of flow simulation studies using irradiated fuel (SCARABEE N program) and for LMFBR transient overpower studies using both fresh and irradiated fuel (CABRI program)

Bruce A units 1 and 2 restart project

International Nuclear Information System (INIS)

Routledge, K.

2006-01-01

This presentation provides an overview of the Bruce A Units 1 and 2 Restart project from the vantage point of the Project Management Contractor (PMC). The presentation will highlight the unique structure of the project, which has been designed to maximize project efficiencies while minimizing the impact to the Bruce Power operational reactors. Efficiency improvements covered in the presentation includes: support services provided to the direct work contractors, radiation protection, worker protection, engineering, field execution, maintenance and facilities. The presentation focusses on the roles of the PMC in helping to ensure the successful outcome of this ambitious reactor refurbishment project. In addition, the Construction Island concept that has been implemented on the project will be presented, with some of the innovative thinking that has gone into its creation. The organization of the PMC and an overview of the project schedule is also presented. AMEC NCL is a privately held consultancy in the Canadian nuclear industry which provides experienced and flexible multi-disciplined resources to support full project management, engineering solutions and safety consultancy services throughout the life cycle of nuclear facilities in Canada, and for customers in related markets in North America and overseas. AMEC NCL is a wholly-owned subsidiary of AMEC plc

Bruce A restart (execution and lessons-learned)

International Nuclear Information System (INIS)

Soini, J.

2011-01-01

Lessons learned with the Bruce Units 3 and 4 restart have been incorporated into the current refurbishment of Units 1 and 2. In addition, lessons learned on the lead unit (U2) are aggressively applied on the lagging unit (U1) to maximize efficiency and productivity. There will be a discussion on how this internal OPEX, along with external lessons learned, are used to continuously improve all aspects of the Bruce A Restart project management cycle, from scope selection, through planning and scheduling, to execution.

Thick-Restart Lanczos Method for Electronic Structure Calculations

International Nuclear Information System (INIS)

Simon, Horst D.; Wang, L.-W.; Wu, Kesheng

1999-01-01

This paper describes two recent innovations related to the classic Lanczos method for eigenvalue problems, namely the thick-restart technique and dynamic restarting schemes. Combining these two new techniques we are able to implement an efficient eigenvalue problem solver. This paper will demonstrate its effectiveness on one particular class of problems for which this method is well suited: linear eigenvalue problems generated from non-self-consistent electronic structure calculations

Present status of Japan materials testing reactor

Energy Technology Data Exchange (ETDEWEB)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Present status of Japan materials testing reactor

International Nuclear Information System (INIS)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

2012-01-01

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

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

Waxy crude oil flow restart ability

Energy Technology Data Exchange (ETDEWEB)

Sierra, Andre Gaona; Varges, Priscilla Ribeiro; Mendes, Paulo Roberto de Souza [Dept. of Mechanical Engineering. Pontificia Universidade Catolica do Rio de Janeiro, RJ (Brazil)], e-mails: prvarges@puc-rio.br, pmendes@puc-rio.br; Ziglio, Claudio [PETROBRAS S.A, R.J., Rio de Janeiro, RJ (Brazil)], e-mail: ziglio@petrobras.com.br

2010-07-01

Under the hot reservoir conditions, waxy crudes behave like Newtonian fluids but once they experience very cold temperatures on the sea floor, the heavy paraffin's begin to precipitate from the solution impacting non- Newtonian flow behavior to the crude (Chang 2000, Lee 2009, Davidson 2004) and begin to deposit on the pipe wall leave blocked of pipeline. This gel cannot be broken with the original steady state flow operating pressure applied before gelation (Chang 1998). Restarting waxy crude oil flows in pipelines is a difficult issue because of the complex rheological behavior of the gelled oil. Indeed, below the WAT, the gelled oil exhibits viscoplastic, thixotropic, temperature-dependent, and compressible properties due to the interlocking gel-like structure formed by the crystallized paraffin compounds and the thermal shrinkage of the oil. The main objective of this work is to determine the minimal pressure to restart the flow, and the relationship between the fluid rheology , pipe geometry and the restart pressure of the flow. Experiments will be performed to investigate the displacement of carbopol aqueous solutions (viscoplastic fluid without thixotropic effects) by Newtonian oil flowing through a strait pipe to validate the experimental apparatus. Therefore, tests will be made with different fluids, like Laponite and waxy crude oils. (author)

Operational Readiness Review Final Report For F-Canyon Restart. Phase 1

Energy Technology Data Exchange (ETDEWEB)

McFarlane, A.F.; Spangler, J.B.

1995-04-05

An independent WSRC Operational Readiness Review was performed for the restart of Phase 1 processing in F-Canyon, Building 221-F. Readiness to restart the Second Plutonium Cycle process and solvent recovery was assessed. The ORR was conducted by an ORR board of ten members with the support of a subject matter expert. The chairman and four members were drawn from the Operational Safety Evaluation Department, ESH& QA Division; additional members were drawn from other WSRC divisions, independent of the F-Canyon operating division (NMPD). Based on the results of the readiness verification assessments performed according to the ORR plan and the validation of pre-restart corrective actions, the WSRC independent ORR Board has concluded that the facility has achieved the state of readiness committed to in the Restart Plan. Also, based on the scope of the ORR, it is the opinion of the board that F-Canyon Phase 1 processes can be restarted without undue risk to the safety of the public and onsite workers and without undue risk to the environment.

Operational Readiness Review Final Report For F-Canyon Restart. Phase 1

International Nuclear Information System (INIS)

McFarlane, A.F.; Spangler, J.B.

1995-01-01

An independent WSRC Operational Readiness Review was performed for the restart of Phase 1 processing in F-Canyon, Building 221-F. Readiness to restart the Second Plutonium Cycle process and solvent recovery was assessed. The ORR was conducted by an ORR board of ten members with the support of a subject matter expert. The chairman and four members were drawn from the Operational Safety Evaluation Department, ESH ampersand QA Division; additional members were drawn from other WSRC divisions, independent of the F-Canyon operating division (NMPD). Based on the results of the readiness verification assessments performed according to the ORR plan and the validation of pre-restart corrective actions, the WSRC independent ORR Board has concluded that the facility has achieved the state of readiness committed to in the Restart Plan. Also, based on the scope of the ORR, it is the opinion of the board that F-Canyon Phase 1 processes can be restarted without undue risk to the safety of the public and onsite workers and without undue risk to the environment

Krylov-Schur-Type restarts for the two-sided arnoldi method

NARCIS (Netherlands)

Zwaan, I.N.; Hochstenbach, M.E.

2017-01-01

We consider the two-sided Arnoldi method and propose a two-sided Krylov-Schurtype restarting method. We discuss the restart for standard Rayleigh-Ritz extraction as well as harmonic Rayleigh-Ritz extraction. Additionally, we provide error bounds for Ritz values and Ritz vectors in the context of

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Reactor BR2: Introduction

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

Some particular aspects of control in nuclear power reactors; Conception de la surete en france et influence des imperatifs de surete sur la conception des reacteurs

Energy Technology Data Exchange (ETDEWEB)

Vathaire, F de; Vernier, Ph; Pascouet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [French] La presente communication propose une synthese de l'experience acquise en France en matiere de surete des reacteurs. Les reacteurs de la filiere graphite-gaz faisant l'objet d'une communication particuliere, on examine ici la surete des autres types de reacteurs etudies en France: - reacteurs eau lourde-gaz, - reacteurs a neutrons rapides, - reacteurs de recherche a eau des types piscines et tank. Les imperatifs de surete propres aux differentes filieres sont developpes, en mettant l'accent sur leur influence sur la conception des reacteurs et sur les limitations de puissance qu'ils entrainent. Les etudes de surete correspondantes sont presentees, en insistant plus particulierement sur les travaux originaux developpes dans ces domaines. On indique notamment les moyens d'essais qui ont ete construits pour ces etudes: le reacteur CABRI, boucle en pile pour essais de depressurisation, boucles hors pile, maquettes, etc. (auteurs)

European in-pile investigations on fast breeder reactor safety

International Nuclear Information System (INIS)

Bailly, J.; Penet, F.; Teague, H.J.

1977-01-01

Because of the difficulties faced by the various organisations responsible for the design and safety analysis of fast reactors, in the conception and execution of perfectly representative experiments corresponding to the various hypothetical accidents considered, the European experts in this field have chosen to adopt a more fundamental approach. By the coordination of the efforts in the different countries and pooling the results, it appears that sufficient data can be assembled to permit them to: isolate the influence of the principle parameters on the course of an accident and to have an appreciation of the possibilities for detection and propagation; and validate calculational models of the essential phenomena, and determine the values of the adjustable parameters used in accident codes. The present paper does not propose to give an exhaustive description of the European experimental programmes, nor of the possibilities for exploiting the results in safety analysis. The aim is simply to show with the aid of results already obtained or expected from the immediate programme, that it should be possible to answer certain fundamental questions concerning those accidents at present considered most important. Particular consideration is given to local blockages (SCARABEE, DFR and MOL 7C programmes), unprotected loss of pumps leading to cooling failure, and inlet blockage (SCARABEE and CABRI), and reactivity excursions (CABRI and VIPER). In addition it is planned to study, as part of the SCARABEE programme, the long-term post-accident behaviour of fuel in some simple configurations. It is shown that these various results should lead to a great improvement in our understanding of the progress of the hypothetical accidents taken into account in the design of fast breeder reactors [fr

LHC Report: Restart preparations continue

CERN Multimedia

Katy Foraz for the LHC team and Julia Trummer for the RP Group

2012-01-01

Maintenance and consolidation work has been progressing well in both the machine and the experiments in preparation for the March restart.   A sample material is attached to the LHC (the white bag taped to the green line), to measure the radiation doses. Additional work was required around Point 5 due to the discovery and repair of a problem with the RF fingers at the connection of two beam vacuum chambers in CMS. The repair has been completed successfully and the sector is now under vacuum. In order to avoid rushing the delicate final operations required for closing the detector, the restart of the machine has been postponed by one week, from 7 March to 14 March. In the machine, the first cool-down to 1.9 K has started in several sectors ,and the cool-down of the whole machine is still planned to be finished by 21 February. The time window between 22 February and 14 March will be dedicated to powering and cryogenic tests. Since 12 December, the Radiation Protection (RP) group has been deep...

LHC Report: Rocky re-start

CERN Multimedia

Barbara Holzer for the LHC Team

2012-01-01

A rocky re-start with beam followed a successful machine development period and the first technical stop of 2012. Today, Friday 11 May, the machine began running again with 1380 bunches.   A short, two-day machine development period was successfully completed on 21-22 April. It focused on topics relevant for the 2012 physics beam operation. This was then followed by a five-day technical stop, the first of the year. The technical stop finished on time on Friday 26 April. The re-start with beam was somewhat tortuous and hampered by an unlucky succession of technical faults leading to extended periods of downtime. The planned intensity increase was put on hold for three days with the machine operating with 1092 bunches and a moderate bunch intensity of 1.3x1011 protons. This delivered a reasonable peak luminosity of 3.6x1033 cm-2s-1 to ATLAS and CMS. Higher than usual beam losses were observed in the ramp and squeeze, and time was required to investigate the causes and to implement mitigati...

The dynamic storage and restart facilities in MABEL-2

International Nuclear Information System (INIS)

Nye, M.T.S.

1983-12-01

MABEL-2 is a FORTRAN program for calculating clad ballooning in a PWR during a LOCA. Originally written with fixed array storage, the use of the code has been extended by including dynamic storage. The lengths of the arrays in the program are set at execution time, varying from run to run. This allows much greater freedom in the choice of mesh and the size of case run. The use of computer memory is also more efficient. In addition a restart facility has been included which allows the user to break off and restart execution of the program (once or many times) during a transient. By using this facility much longer calculations can be run. Should an error in either input data or program become apparent late in a transient, the case need only be re-run from the last dump because some input data can be altered at restart. The use of these new facilities and the coding changes are described. (author)

Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

International Nuclear Information System (INIS)

Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

1994-01-01

Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

International Nuclear Information System (INIS)

Easterling, T.C.; Hightower, N.T.; Smith, D.C.; Amos, C.N.

1992-01-01

An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs

Environmental Information Document: L-reactor reactivation

International Nuclear Information System (INIS)

Mackey, H.E. Jr.

1982-04-01

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program

Environmental Information Document: L-reactor reactivation

Energy Technology Data Exchange (ETDEWEB)

Mackey, H.E. Jr. (comp.)

1982-04-01

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

Effect of hydraulic retention time on deterioration/restarting of sludge anaerobic digestion: Extracellular polymeric substances and microbial response.

Science.gov (United States)

Wei, Liangliang; An, Xiaoyan; Wang, Sheng; Xue, Chonghua; Jiang, Junqiu; Zhao, Qingliang; Kabutey, Felix Tetteh; Wang, Kun

2017-11-01

In this study, the transformation of the sludge-related extracellular polymeric substances (EPS) during mesophilic anaerobic digestion was characterized to assess the effect of hydraulic retention time (HRT) on reactor deterioration/restarting. Experimental HRT variations from 20 to 15 and 10d was implemented for deterioration, and from 10 to 20d for restarting. Long-term digestion at the lowest HRT (10d) resulted in significant accumulation of hydrolyzed hydrophobic materials and volatile fatty acids in the supernatants. Moreover, less efficient hydrolysis of sludge EPS, especially of proteins related substances which contributed to the deterioration of digester. Aceticlastic species of Methanosaetaceae decreased from 36.3% to 27.6% with decreasing HRT (20-10d), while hydrogenotrophic methanogens (Methanomicrobiales and Methanobacteriales) increased from 30.4% to 38.3%. Proteins and soluble microbial byproducts related fluorophores in feed sludge for the anaerobic digester changed insignificantly at high HRT, whereas the fluorescent intensity of fulvic acid-like components declined sharply once the digestion deteriorated. Copyright © 2017 Elsevier Ltd. All rights reserved.

Annual report of the Neutron Irradiation and Testing Reactor Center. FY2007. April 1, 2007 - March 31, 2008

International Nuclear Information System (INIS)

2009-03-01

The Japan Materials Testing Reactor (JMTR), achieving first criticality in March 1968, has been used in testing the durability and integrity of reactor fuels and components, basic nuclear research, the production of radioisotopes (RIs), and other purposes. The JMTR, however, stopped in August 2006 after its 165th operation cycle, and is currently under going partial renewal of reactor facilities and installation of new irradiation Facilities, geared toward being restarted in 2011. In addition, to cope with the strong requests from users to improve usability of the JMTR, efforts are being made to increase reactor operation efficiency, shorten the turnaround time for obtaining results, and other necessary tasks for JMTR to commence reoperation. The present report summarizes the activities carried out in 2007 for the refurbishment and restart of JMTR. (author)

A review of fast reactor activities in Italy

International Nuclear Information System (INIS)

Tavoni, R.

1996-01-01

In this paper, Italian activities on liquid metal fast reactors are shown for the period May 1995 - April 1996. During this period the ENEA collaboration with General Electric on ALMR came to an end as a consequence of the reduced effort on the design development. Nevertheless ENEA completed the studies on the PRISM Mod B oxide burner core, the neutronic configuration of which was presented at last year's meeting. Some results of the dynamic calculations are shown. ENEA participated in the IAEA/EC benchmark on the comparative calculations for severe accident in BN-800 reactor. A complete neutronic evaluation has been made including power distribution, Doppler, sodium void and material coefficients. Activities on seismic isolation are also outlined. The Italian contribution to SPX restart and operation is described; some information about the complementary convention of the Nersa society is given, together with the Italian industry participation in the SPX restart. (author)

Lessons learned form high-flux isotope reactor restart efforts

International Nuclear Information System (INIS)

Dahl, T.L.

1989-01-01

When the high-flux isotope reactor's (HFIR's) pressure vessel irradiation surveillance specimens were examined in December 1986, unexpected embrittlement was found. The resulting investigation disclosed widespread deficiencies in quality assurance and management practices. On March 24, 1987, the US Department of Energy (DOE) mandated a shutdown of all five Oak Ridge National Laboratory (ORNL) research reactors. Since the beginning of 1987, 18 different formal review groups have evaluated the management and operations of the HFIR. The root cause of the identified deficiencies in the HFIR program was defined as a lack of rigor in management practices and complacency built on twenty years of trouble-free operation. A number of lessons can be learned from the HFIR experience. Particular insight can be gained by comparing the HFIR organization prior to the shutdown with the organization that exists today. Key elements in such a comparison include staffing, funding, discipline, and formality in operations, maintenance, and management

RELAP5-3D Resolution of Known Restart/Backup Issues

Energy Technology Data Exchange (ETDEWEB)

Mesina, George L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Anderson, Nolan A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-12-01

The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequential verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.

Checkpoint-dependent RNR induction promotes fork restart after replicative stress.

Science.gov (United States)

Morafraile, Esther C; Diffley, John F X; Tercero, José Antonio; Segurado, Mónica

2015-01-20

The checkpoint kinase Rad53 is crucial to regulate DNA replication in the presence of replicative stress. Under conditions that interfere with the progression of replication forks, Rad53 prevents Exo1-dependent fork degradation. However, although EXO1 deletion avoids fork degradation in rad53 mutants, it does not suppress their sensitivity to the ribonucleotide reductase (RNR) inhibitor hydroxyurea (HU). In this case, the inability to restart stalled forks is likely to account for the lethality of rad53 mutant cells after replication blocks. Here we show that Rad53 regulates replication restart through the checkpoint-dependent transcriptional response, and more specifically, through RNR induction. Thus, in addition to preventing fork degradation, Rad53 prevents cell death in the presence of HU by regulating RNR-expression and localization. When RNR is induced in the absence of Exo1 and RNR negative regulators, cell viability of rad53 mutants treated with HU is increased and the ability of replication forks to restart after replicative stress is restored.

ESP – Data from Restarted Life Tests of Various Silicon Materials

Energy Technology Data Exchange (ETDEWEB)

Schneider, Jim

2010-10-06

Current funding has allowed the restart of testing of various silicone materials placed in Life Tests or Aging Studies from past efforts. Some of these materials have been in test since 1982, with no testing for approximately 10 years, until funding allowed the restart in FY97. Charts for the various materials at different thickness, compression, and temperature combinations illustrate trends for the load-bearing properties of the materials.

Legal claims against Belgian reactors?; Rechtsmittel gegen belgische Reaktoren?

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

The Belgian reactors Tihange 2 and Doel 3 have been restarted in November 2015 after the problem of hydrogen flakes in the reactor pressure vessels had been investigated. The permission to restart has been the object both of critical statements by the German Federal Ministry of the Environment (BMUB) and of lawsuits filed with Belgian law courts by a group of German municipalities led by the city of Aachen and by the Land North-Rhine-Westphalia. According to a general principle of the law of nations, a state is not permitted to operate installations near its border, which cause significant environmental damage in a neighbouring state. However, it is not quite clear how this principle applies to the issue of potential accidents of nuclear power plants. According to the author, a tangible threat of an accident is required; mere doubts and concerns about the extent of safety margins are not sufficient.

Economic targets for small PWR reactor designs

International Nuclear Information System (INIS)

Board, J.

1991-01-01

Small reactors are likely to be less economic than large reactors, but the lower financial exposure with small reactors may be attractive to utilities contemplating a restart to a nuclear programme. New nuclear plant can be economic, but success will depend more on how the plant are built, rather than what type or size is built. A target for new plant for operation early in the next century should be a generation cost of 3p to 3.5 p/kWh. This corresponds to an overnight capital cost of Pound 1000/kWh to Pound 1100/kWh. (author)

Failure Recovery via RESTART: Wallclock Models

DEFF Research Database (Denmark)

Asmussen, Søren; Rønn-Nielsen, Anders

A task such as the execution of a computer program or the transfer of a file on a communications link may fail and then needs to be restarted. Let the ideal task time be a constant $\\ell$ and the actual task time $X$, a random variable. Tail asymptotics for $\\mathbb{P}(X>x)$ is given under three ...

LHC Experiments: refinements for the restart

CERN Multimedia

2009-01-01

As the LHC restart draws closer, the Bulletin will be taking a look at how the six LHC experiments are preparing and what they have been up to since last September. In this issue we start with a roundup of the past 10 months of activity at CMS and ATLAS, both technical work and outreach activities.

Beleaguered LHC gears up for restart

CERN Multimedia

Cartwright, Jon

2009-01-01

"The Large Hadron Collider (LHC) is finally set to restart in mid-November following last year's accident. Initially it will collide protons at an energy of only 3.5 TeV per beam, and staff at Cern will have to wait until late next year before trying to run the collider at its maximum energy" (0.75 page)

RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

1981-12-01

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

Record of Decision; Continued operation of K, L, and P Reactors, Savannah River Site, Aiken, South Carolina

International Nuclear Information System (INIS)

1991-01-01

The US Department of Energy (DOE) has considered the environmental impacts, benefits and costs, and institutional and programmatic needs associated with continued operation of the Savannah River Site (SRS) reactors, and has decided that it will continue to operate K and L Reactors at SRS, and will terminate operation of P Reactor in the immediate future and maintain it in cold standby. For P Reactor, this will involve the reactor's defueling; storage of its heavy water moderator in tanks in the reactor building; shutdown of reactor equipment and systems in a protected condition to prevent deterioration; and maintenance of the reactor in a defueled, protected status by a skeleton staff, which would permit any future decision to refuel and restart. Currently committed and planned upgrade activities will be discontinued for P Reactor. DOE will proceed with the safety upgrades and management system improvements currently scheduled for K Reactor in its program to satisfy the criteria of the Safety Evaluation Report (SER), and will conduct an Operational Readiness Review (ORR). The satisfaction of the SER criteria and completion of the ORR will demonstrate that the safety and health criteria for the resumption of production have been met. Reactor restart is expected to be in the third quarter of 1991 for K Reactor

Stuck fermentation: development of a synthetic stuck wine and study of a restart procedure.

Science.gov (United States)

Maisonnave, Pierre; Sanchez, Isabelle; Moine, Virginie; Dequin, Sylvie; Galeote, Virginie

2013-05-15

Stuck fermentation is a major problem in winemaking, resulting in large losses in the wine industry. Specific starter yeasts are used to restart stuck fermentations in conditions determined essentially on the basis of empirical know-how. We have developed a model synthetic stuck wine and an industrial process-based procedure for restarting fermentations, for studies of the conditions required to restart stuck fermentations. We used a basic medium containing 13.5% v/v ethanol and 16 g/L fructose, pH 3.3, to test the effect of various nutrients (vitamins, amino acids, minerals, oligoelements), with the aim of developing a representative and discriminative stuck fermentation model. Cell growth appeared to be a key factor for the efficient restarting of stuck fermentations. Micronutrients, such as vitamins, also strongly affected the efficiency of the restart procedure. For the validation of this medium, we compared the performances of three wine yeast strains in the synthetic stuck fermentation and three naturally stuck wine fermentations. Strain performance was ranked similar in the synthetic medium and in the "Malbec" and "Sauvignon" natural stuck wines. However, two strains were ranked differently in the "Gros Manseng" stuck wine. Nutrient content seemed to be a crucial factor in fermentation restart conditions, generating differences between yeast strains. However, the specific sensitivity of yeast strains to the composition of the wine may also have had an effect. Copyright © 2013 Elsevier B.V. All rights reserved.

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

Annual report of the Neutron Irradiation and Testing Reactor Center. FY 2007. April 1, 2007 - March 31, 2008

International Nuclear Information System (INIS)

2009-03-01

The Japan Materials Testing Reactor (JMTR), achieving the first criticality in March 1968, has been used to test the durability and integrity of reactor fuels and components, basic nuclear research, production of radioisotopes (RIs), and other purposes. The JMTR, however, was halted in August 2006 after its 165th cycle operation, and is currently undergoing partial renewal of the apparatus and installation of new irradiation equipment, aiming at restarting from 2011. In addition, to cope with strong requests from users to improve the usability of the JMTR, efforts are being made to increase reactor operating efficiency, shorten the turnaround time for obtaining results, and conduct other necessary tasks for the JMTR to recommence reoperation. The present report summarizes the activities carried out in 2007 for the refurbishment and restart of the JMTR. (author)

Westinghouse independent safety review of Savannah River production reactors

International Nuclear Information System (INIS)

Leggett, W.D.; McShane, W.J.; Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E.; Call, D.W.

1989-01-01

Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K, L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours ampersand Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours ampersand Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone. 37 refs., 1 fig., 3 tabs

Westinghouse independent safety review of Savannah River production reactors

Energy Technology Data Exchange (ETDEWEB)

Leggett, W.D.; McShane, W.J. (Westinghouse Hanford Co., Richland, WA (USA)); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.); Toto, G. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Services Div.); Fauske, H.K. (Fauske and Associates, Inc., Burr Ridge, IL (USA)); Call, D.W. (Westinghouse Savannah R

1989-04-01

Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

Restart of the LHC in 2009

CERN Multimedia

Corinne Pralavorio

The restart of the LHC during the summer 2009 has been confirmed today, the 5 December. An updated report on the incident which damaged sector 3-4 has just been published. It gives details on the damage caused by the incident and explains the ongoing repairs and the new systems being put into place to reinforce the safety of the machine. Click here to see the report.

L-Reactor Habitat Mitigation Study

International Nuclear Information System (INIS)

1988-02-01

The L-Reactor Fish and Wildlife Resource Mitigation Study was conducted to quantify the effects on habitat of the L-Reactor restart and to identify the appropriate mitigation for these impacts. The completed project evaluated in this study includes construction of a 1000 acre reactor cooling reservoir formed by damming Steel Creek. Habitat impacts identified include a loss of approximately 3,700 average annual habitat units. This report presents a mitigation plan, Plan A, to offset these habitat losses. Plan A will offset losses for all species studied, except whitetailed deer. The South Carolina Wildlife and Marine Resources Department strongly recommends creation of a game management area to provide realistic mitigation for loss of deer habitats. 10 refs., 5 figs., 3 tabs

Public communication toward Monju restart

International Nuclear Information System (INIS)

Aoki, Tadao

2001-01-01

Five years have gone by since the sodium leak took place at a prototype FBR Monju. Looking back upon that time, one journalist said, The Monju accident was technically far from the serious one as being reported in the media. Had it not been for the infamous 'accident cover-up', an uproar must have calmed down in a month. But an unexpectedly large negative public reaction has kept Monju idle all these years. What had really happened? There was a false report on the time of first entry to the piping room or the sodium-leak spot. Contrary to the fact that five staffs did enter the room at 2:00 am, PNC failed to mention it at a first press conference held at 8:30 am. Instead, PNC created a fictitious time of entry at 10:00 am and reported it to the authorities in a formal document. Another mishap was a video cover-up operation. A year and three months later, an explosion accident took place at PNC's Tokai Reprocessing Facility and similar mishap was repeated then, causing a fatal damage to the PNC's reputation. Public opinion polls taken by mass media have concluded that PNC is 'bureaucratic, closed, slow in coping with situation and untrustworthy'. PNC struggle began - struggle to regain public trust. A series of mishaps at PNC have created an anxiety and distrust about nuclear energy among the nation. In order to restore the trust of the nation, STA, a government agency supervising PNC, decided that PNC be reorganised to make a new start as Japan Nuclear Cycle Development Institute (JNC) on October 1, 1998. In the start of the new organisation, JNC is expected to carry out operations placing priority on the locality of its facilities. The most precious lesson learned from the Monju accident is the importance of public communication. Currently undertaking activities toward Monju restart are; 1) public opinion monitoring, 2) social meetings, 3) strengthening publicity activities, 4) dialogue with local administration officials and opinion leaders, 5) 'open meeting

Interpretation of the CABRI LT1 test with SAS4A-code analysis

International Nuclear Information System (INIS)

Sato, Ikken; Onoda, Yu-uichi

2001-03-01

In the CABRI-FAST LT1 test, simulating a ULOF (Unprotected Loss of Flow) accident of LMFBR, pin failure took place rather early during the transient. No fuel melting is expected at this failure because the energy injection was too low and a rapid gas-release-like response leading to coolant-channel voiding was observed. This channel voiding was followed by a gradual fuel breakup and axial relocation. With an aid of SAS4A analysis, interpretation of this test was performed. Although the original SAS4A model was not well fitted to this type of early pin failure, the global behavior after the pin failure was reasonably simulated with temporary modifications. Through this study, gas release behavior from the failed fuel pin and its effect on further transient were well understood. It was also demonstrated that the SAS4A code has a potential to simulate the post-failure behavior initiated by a very early pin failure provided that necessary model modification is given. (author)

A review of the Indian fast reactor programme

International Nuclear Information System (INIS)

Chetal, S.C.

1989-01-01

Fast Breeder Test Reactor (FBTR) in India is ready for restart. Satisfactory progress has been made in the design of Prototype Fast Breeder Reactor (PFBR). Conceptual design work for the important systems and components has been completed. Cost estimation is in progress. Detailed project report for the financial sanction is under completion stage and is planned to be submitted to the Government this year. Draft Safety criteria prepared by a sub-committee on behalf of the Regulatory Board have been discussed and will be issued shortly. (author)

Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

International Nuclear Information System (INIS)

Roudier, S.; Badel, D.; Beraha, R.; Champ, M.; Tricot, N.; Tran Dai, P.

1994-01-01

The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: 1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; 2) guidelines for nuclear design and manufacturing requirements related to safety and 3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs

Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

Energy Technology Data Exchange (ETDEWEB)

Roudier, S [Direction de la Surete des Installations Nucleaires, Fontenay-aux-Roses (France); Badel, D; Beraha, R [Direction Regionale de l` Industrie, de la Recherche et de l` Environnement Rhone-Alpes, Lyon (France); Champ, M; Tricot, N; Tran Dai, P [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

1994-12-31

The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: (1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; (2) guidelines for nuclear design and manufacturing requirements related to safety and (3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs.

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

Sobre el estudio de la función cuadrática y su relación con el área de algunas figuras y su visualización usando Cabri II Plus

OpenAIRE

Vera Barrios, Dídimo

2013-01-01

Propuesta didáctica en donde se presentan algunas actividades dinámicas usando Cabri II Plus con el fin de relacionar algunos aspectos que tienen que ver con el área de algunas figuras y la función cuadrática.

The IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

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

Extending reactor time-to-poison and reducing poison shutdown time by pre-shutdown power alterations

Energy Technology Data Exchange (ETDEWEB)

Kerr, Edward

1963-10-15

Manipulation of reactor power prior to shutdown and increasing the time- to-poison a sufficient amount to enable the required maintenance work to be completed and the reactor immediately restarted are discussed. The method employed in the NRU Reactor to gain the maximum timeto-poison with the least production loss is outlined. The method is based on intuition and is described by means of an analog of the iodine--xenon equations rather than the equations themselves. (C.E.S.)

Challenges of restarting Bruce Units 3 and 4 from a chemistry and materials perspective

International Nuclear Information System (INIS)

Roberts, J.G.; Langguth, K.

2005-01-01

In 2001, Bruce Power leased the Bruce Units 1-8 reactors from Ontario Power Generation. Bruce Power decided to restart Bruce Units 3 and 4 following a condition assessment of Bruce A Units 3 and 4. This paper describes the challenges that were encountered and how they were overcome, specifically for heat transport system chemistry in order to adequately protect carbon steel surfaces. The heat transport system, by design, has close inter-relations with other station systems and the related issues of some of these systems are also discussed. Considerations of material impacts have significant influences on the approach to, and control of, chemistry. Specific material impacts led to a novel, and successful, approach. This approach was arrived at following significant efforts by a multi-disciplinary team of operations, maintenance and chemistry staff. The issues, approaches considered and solutions used for a successful outcome will be presented. (author)

Challenges of restarting Bruce Units 3 and 4 from a chemistry and materials perspective

Energy Technology Data Exchange (ETDEWEB)

Roberts, J.G.; Langguth, K. [Bruce Power, Tiverton, Ontario (Canada)

2005-07-01

In 2001, Bruce Power leased the Bruce Units 1-8 reactors from Ontario Power Generation. Bruce Power decided to restart Bruce Units 3 and 4 following a condition assessment of Bruce A Units 3 and 4. This paper describes the challenges that were encountered and how they were overcome, specifically for heat transport system chemistry in order to adequately protect carbon steel surfaces. The heat transport system, by design, has close inter-relations with other station systems and the related issues of some of these systems are also discussed. Considerations of material impacts have significant influences on the approach to, and control of, chemistry. Specific material impacts led to a novel, and successful, approach. This approach was arrived at following significant efforts by a multi-disciplinary team of operations, maintenance and chemistry staff. The issues, approaches considered and solutions used for a successful outcome will be presented. (author)

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

The Efficacy of a Restart Break for Recycling with Optimal Performance Depends Critically on Circadian Timing

Science.gov (United States)

Van Dongen, Hans P.A.; Belenky, Gregory; Vila, Bryan J.

2011-01-01

Objectives: Under simulated shift-work conditions, we investigated the efficacy of a restart break for maintaining neurobehavioral functioning across consecutive duty cycles, as a function of the circadian timing of the duty periods. Design: As part of a 14-day experiment, subjects underwent two cycles of five simulated daytime or nighttime duty days, separated by a 34-hour restart break. Cognitive functioning and high-fidelity driving simulator performance were tested 4 times per day during the two duty cycles. Lapses on a psychomotor vigilance test (PVT) served as the primary outcome variable. Selected sleep periods were recorded polysomnographically. Setting: The experiment was conducted under standardized, controlled laboratory conditions with continuous monitoring. Participants: Twenty-seven healthy adults (13 men, 14 women; aged 22–39 years) participated in the study. Interventions: Subjects were randomly assigned to a nighttime duty (experimental) condition or a daytime duty (control) condition. The efficacy of the 34-hour restart break for maintaining neurobehavioral functioning from the pre-restart duty cycle to the post-restart duty cycle was compared between these two conditions. Results: Relative to the daytime duty condition, the nighttime duty condition was associated with reduced amounts of sleep, whereas sleep latencies were shortened and slow-wave sleep appeared to be conserved. Neurobehavioral performance measures ranging from lapses of attention on the PVT to calculated fuel consumption on the driving simulators remained optimal across time of day in the daytime duty schedule, but degraded across time of night in the nighttime duty schedule. The 34-hour restart break was efficacious for maintaining PVT performance and other objective neurobehavioral functioning profiles from one duty cycle to the next in the daytime duty condition, but not in the nighttime duty condition. Subjective sleepiness did not reliably track objective neurobehavioral

Berkeley lab checkpoint/restart (BLCR) for Linux clusters

International Nuclear Information System (INIS)

Hargrove, Paul H; Duell, Jason C

2006-01-01

This article describes the motivation, design and implementation of Berkeley Lab Checkpoint/Restart (BLCR), a system-level checkpoint/restart implementation for Linux clusters that targets the space of typical High Performance Computing applications, including MPI. Application-level solutions, including both checkpointing and fault-tolerant algorithms, are recognized as more time and space efficient than system-level checkpoints, which cannot make use of any application-specific knowledge. However, system-level checkpointing allows for preemption, making it suitable for responding to ''fault precursors'' (for instance, elevated error rates from ECC memory or network CRCs, or elevated temperature from sensors). Preemption can also increase the efficiency of batch scheduling; for instance reducing idle cycles (by allowing for shutdown without any queue draining period or reallocation of resources to eliminate idle nodes when better fitting jobs are queued), and reducing the average queued time (by limiting large jobs to running during off-peak hours, without the need to limit the length of such jobs). Each of these potential uses makes BLCR a valuable tool for efficient resource management in Linux clusters

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1988-01-01

According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

Funcionalidad de Juegos de Estrategia Virtuales y del Software Cabri-géomètre II en el Aprendizaje de la Simetría en Secundaria

Directory of Open Access Journals (Sweden)

Guadalupe Rodríguez

2010-06-01

Full Text Available Presentamos resultados de un estudio exploratorio cuyo propósito es indagar sobre el uso del Cabri-Géomètre II y de un juego matemático virtual de estrategia cuando se incluyen en clases ordinarias de matemáticas de estudiantes de 12-13 años en una escuela secundaria pública de México. Para este trabajo, el tema abordado fue la simetría. El análisis de datos permitió ver la funcionalidad de estos ambientes para transformar las nociones del alumno sobre el tema y para reconocer que se puede avanzar en el desarrollo de un pensamiento matemático distinto a través de ambientes de aprendizaje computacionales. We present the results of an exploratory study whose purpose was to explore the use by 12-13 years old students of Cabri-Géomètre II and a mathematical strategic virtual game in a Mexican state school. For this paper, the mathematical focus was on symmetry. The data analysis allowed us to observe the functionality of this learning environment to transform the students’ conceptions and to recognize that a different type of mathematical thinking can be developed with these computational learning environments.

Friction Stir Weld Restart+Reweld Repair Allowables

Science.gov (United States)

Clifton, Andrew

2008-01-01

A friction stir weld (FSW) repair method has been developed and successfully implemented on Al 2195 plate material for the Space Shuttle External Fuel Tank (ET). The method includes restarting the friction stir weld in the termination hole of the original weld followed by two reweld passes. Room temperature and cryogenic temperature mechanical properties exceeded minimum FSW design strength and compared well with the development data. Simulated service test results also compared closely to historical data for initial FSW, confirming no change to the critical flaw size or inspection requirements for the repaired weld. Testing of VPPA fusion/FSW intersection weld specimens exhibited acceptable strength and exceeded the minimum design value. Porosity, when present at the intersection was on the root side toe of the fusion weld, the "worst case" being 0.7 inch long. While such porosity may be removed by sanding, this "worst case" porosity condition was tested "as is" and demonstrated that porosity did not negatively affect the strength of the intersection weld. Large, 15-inch "wide panels" FSW repair welds were tested to demonstrate strength and evaluate residual stresses using photo stress analysis. All results exceeded design minimums, and photo stress analysis showed no significant stress gradients due to the presence of the restart and multi-pass FSW repair weld.

Final Report for the Restart of the Waste Characterization, Reduction and Repackaging Facility (WCRRF) Contractor Readiness Assessment (CRA)

Energy Technology Data Exchange (ETDEWEB)

Stephens, Gregory Mark [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-02-22

The Los Alamos National Laboratory (LANL or Laboratory) Contractor Readiness Assessment (CRA) required for restart of the Technical Area (TA) 50 Waste Characterization, Reduction, and Repackaging Facility (WCRRF) for remediated nitrate salt (RNS) waste operations was performed in compliance with the requirements of Department of Energy (DOE) Order (O) 425.1D, Verification of Readiness to Start Up or Restart Nuclear Facilities, and LANL procedure FSD-115-001, Verification of Readiness to Start Up or Restart LANL Nuclear Facilities, Activities, and Operations.

Activity report of Reactor Physics Division : 1990

International Nuclear Information System (INIS)

Mohanakrishnan, P.

1991-01-01

The major Research and Development and Project activities carried out during the year 1990 in Reactor Physics Division are presented in the form of summaries in this report. The various activities are organised under the following areas : (1) Nuclear Data Evaluation, Processing and Validation, (2) Core Physics and Analysis, (3) Reactor Kinetics and Safety Analysis, (4) Noise Analysis, and (5) Radiation Transport and Shielding. FBTR was restarted in July 1990 and the power was raised upto 500 kW. A number of low power physics experiments on reactivity coefficients, kinetics and noise, neutron flux and gamma dose in B cells, were performed, which are discussed in this report. (author). figs., tabs

Documentation of a restart option for the U.S. Geological Survey coupled Groundwater and Surface-Water Flow (GSFLOW) model

Science.gov (United States)

Regan, R. Steve; Niswonger, Richard G.; Markstrom, Steven L.; Barlow, Paul M.

2015-10-02

A new option to write and read antecedent conditions (also referred to as initial conditions) has been developed for the U.S. Geological Survey (USGS) Groundwater and Surface-Water Flow (GSFLOW) numerical, hydrologic simulation code. GSFLOW is an integration of the USGS Precipitation-Runoff Modeling System (PRMS) and USGS Modular Groundwater-Flow Model (MODFLOW), and provides three simulation modes: MODFLOW-only, PRMS-only, and GSFLOW (or coupled). The new capability, referred to as the restart option, can be used for all three simulation modes, such that the results from a pair (or set) of spin-up and restart simulations are nearly identical to results produced from a continuous simulation for the same time period. The restart option writes all results to files at the end of a spin-up simulation that are required to initialize a subsequent restart simulation. Previous versions of GSFLOW have had some capability to save model results for use as antecedent condiitions in subsequent simulations; however, the existing capabilities were not comprehensive or easy to use. The new restart option supersedes the previous methods. The restart option was developed in collaboration with the National Oceanic and Atmospheric Administration, National Weather Service as part of the Integrated Water Resources Science and Services Partnership. The primary focus for the development of the restart option was to support medium-range (7- to 14-day) forecasts of low streamflow conditions made by the National Weather Service for critical water-supply basins in which groundwater plays an important role.

Results of environmental radiation monitoring and meteorology measurements (material prepared for obtaining the licence for RA reactor experimental operation)

International Nuclear Information System (INIS)

1980-10-01

According to the demands for obtaining the licence for restarting the Ra reactor and the experimental operation this document includes the radiation monitoring measured data in the working space and environment of the RA reactor, i.e. Boris Kidric Institute. The meteorology measured data are included as well. All the measurements are performed according to the radiation protection program applied actually from the first reactor start-up at the end of 1959 [sr

Re-starting an Arnoldi iteration

Energy Technology Data Exchange (ETDEWEB)

Lehoucq, R.B. [Argonne National Lab., IL (United States)

1996-12-31

The Arnoldi iteration is an efficient procedure for approximating a subset of the eigensystem of a large sparse n x n matrix A. The iteration produces a partial orthogonal reduction of A into an upper Hessenberg matrix H{sub m} of order m. The eigenvalues of this small matrix H{sub m} are used to approximate a subset of the eigenvalues of the large matrix A. The eigenvalues of H{sub m} improve as estimates to those of A as m increases. Unfortunately, so does the cost and storage of the reduction. The idea of re-starting the Arnoldi iteration is motivated by the prohibitive cost associated with building a large factorization.

Restart plan for the prototype vertical denitration calciner

Energy Technology Data Exchange (ETDEWEB)

SUTTER, C.S.

1999-09-01

Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing.

Restart plan for the prototype vertical denitration calciner

International Nuclear Information System (INIS)

SUTTER, C.S.

1999-01-01

Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing

Initialization and Restart in Stochastic Local Search: Computing a Most Probable Explanation in Bayesian Networks

Science.gov (United States)

Mengshoel, Ole J.; Wilkins, David C.; Roth, Dan

2010-01-01

For hard computational problems, stochastic local search has proven to be a competitive approach to finding optimal or approximately optimal problem solutions. Two key research questions for stochastic local search algorithms are: Which algorithms are effective for initialization? When should the search process be restarted? In the present work we investigate these research questions in the context of approximate computation of most probable explanations (MPEs) in Bayesian networks (BNs). We introduce a novel approach, based on the Viterbi algorithm, to explanation initialization in BNs. While the Viterbi algorithm works on sequences and trees, our approach works on BNs with arbitrary topologies. We also give a novel formalization of stochastic local search, with focus on initialization and restart, using probability theory and mixture models. Experimentally, we apply our methods to the problem of MPE computation, using a stochastic local search algorithm known as Stochastic Greedy Search. By carefully optimizing both initialization and restart, we reduce the MPE search time for application BNs by several orders of magnitude compared to using uniform at random initialization without restart. On several BNs from applications, the performance of Stochastic Greedy Search is competitive with clique tree clustering, a state-of-the-art exact algorithm used for MPE computation in BNs.

Status of researches in the field of safety of pressurized water reactors

International Nuclear Information System (INIS)

Couturier, Jean; Schwarz, Michel

2017-01-01

This collective publication proposes a synthesis of the status of researches performed in the field of safety of pressurized water reactors. They may discuss past, current and projected research works, involved actors, or lessons learned from these works. The authors propose a presentation of some research tools privileged by the IRSN for these researches: the CABRI and PHEBUS reactors, the GALAXIE experimental platform, and some other installations. Then they address researches related to loss-of-coolant accidents (two-phase thermohydraulics, fuel rod behaviour), to reactivity accidents, to accidents related to dewatering of irradiated fuel storage pools, to fires, to extreme aggressions of natural origin (earthquake, extreme flooding), to core fusion accidents (core heating and fusion within the vessel, vessel failure and apron erosion by corium, containment enclosure dynamic loading, release of radioactive products), to the behaviour of nuclear plant important metallic or civil works components and notably to their ageing, to organisational and human factors or more generally to social and human sciences (design of control rooms, safety organisation and management in EDF nuclear plants), and to other issues and research perspectives

Markov Renewal Methods in Restart Problems in Complex Systems

DEFF Research Database (Denmark)

Asmussen, Søren; Lipsky, Lester; Thompson, Stephen

A task with ideal execution time L such as the execution of a computer program or the transmission of a file on a data link may fail, and the task then needs to be restarted. The task is handled by a complex system with features similar to the ones in classical reliability: failures may...

Restart Testing Program for piping following steam generator replacement at North Anna Unit 1

International Nuclear Information System (INIS)

Bain, R.A.; Bayer, R.K.

1993-01-01

In order to provide assurance that the effects of performing steam generator replacement (SGR) at North Anna unit 1 had no adverse impact on plant piping systems, a cold functional verification restart testing program was developed. This restart testing program was implemented in lieu of a hot functional testing program normally used during the initial startup of a nuclear plant. A review of North Anna plant-specific and generic U.S. Nuclear Regulatory Commission requirements for restart testing was performed to ensure that no mandatory hot functional testing was required. This was determined to be the case, and the development of a cold functional test program was initiated. The cold functional test had inherent advantages as compared to the hot functional testing, while still providing assurance of piping system adequacy. The advantages of the cold verification program included reducing risk to personnel from hot piping, increasing the accuracy of measurements with the improvement in work conditions, eliminating engineering activities during the heatup process, and being able to record measurements as construction work was completed allowing for rework or repair of components if required. To ensure the effectiveness of the cold verification program, a project procedure was generated to identify the personnel, equipment, and measurement requirements. An engineering calculation was issued to document the scope of the restart test program, and an additional calculation was developed to provide acceptance criteria for the critical commodity measurements

The role of internal and external control for mitigating or preventing LMR accidents

International Nuclear Information System (INIS)

Waltar, A.E.; Padilla, A.; Seeman, S.E.

1987-01-01

For the safety assessment of LMFBRs, much effort has been devoted to the analyses of LOF accident sequences with an emphasis on initiating-phase (IP) energetics (LOF-d-TOP event). Important knowledge and experiences on the IP energetics have been accumulated through reactor studies and in-pile experiment analyses, typically for the CABRI experiments. The present paper summarises the current understanding of key phenomenology relevant to the IP energetics based on the CABRI experiment analyses and the validation study for the PAPAS-2S, SAS3D and SAS4A codes. (author)

Re-examining reactor vessel embrittlement at Chooz A

International Nuclear Information System (INIS)

Guilleret, J.-C.

1988-01-01

The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

WRAP: a water reactor analysis package

International Nuclear Information System (INIS)

Anderson, M.M.

1977-06-01

The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

Coder period - amplitude; Codeur periode - amplitude

Energy Technology Data Exchange (ETDEWEB)

Da Costa Viera, D; Mourchou, R; Calvet, A; Pacchiani, L; Tattegrain, A

1966-07-01

For some flowmeters on nuclear reactors, apparatus which deliver an ac signal proportional to the flow, are usually used. This frequency information is converted into a continuous voltage with a coder. The described coder was studied for the Cabri reactor which was requiring a short response delay. (A.L.B.)

Amplifiers with ground-isolated inputs and outputs

International Nuclear Information System (INIS)

Da Costa Vieira, David; Merite, Bernard; Tattegrain, Alain

1969-06-01

The amplifiers described in this note aim at ensuring a connection between different apparatuses with grounds being at a different potential. They will be inserted in the measurement channels of the Cabri reactor

Commentary: restarting NTD programme activities after the Ebola outbreak in Liberia.

Science.gov (United States)

Thomas, Brent C; Kollie, Karsor; Koudou, Benjamin; Mackenzie, Charles

2017-05-01

It is widely known that the recent Ebola Virus Disease (EVD) in West Africa caused a serious disruption to the national health system, with many of ongoing disease focused programmes, such as mass drug administration (MDA) for onchocerciasis (ONC), lymphatic filariasis (LF) and schistosomiasis (SCH), being suspended or scaled-down. As these MDA programmes attempt to restart post-EVD it is important to understand the challenges that may be encountered. This commentary addresses the opinions of the major health sectors involved, as well as those of community members, regarding logistic needs and challenges faced as these important public health programmes consider restarting. There appears to be a strong desire by the communities to resume NTD programme activities, although it is clear that some important challenges remain, the most prominent being those resulting from the severe loss of trained staff.

Renewal of reactor cooling system of JMTR. Reactor building site

International Nuclear Information System (INIS)

Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

2012-03-01

The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

Keeping checkpoint/restart viable for exascale systems.

Energy Technology Data Exchange (ETDEWEB)

Riesen, Rolf E.; Bridges, Patrick G. (IBM Research, Ireland, Mulhuddart, Dublin); Stearley, Jon R.; Laros, James H., III; Oldfield, Ron A.; Arnold, Dorian (University of New Mexico, Albuquerque, NM); Pedretti, Kevin Thomas Tauke; Ferreira, Kurt Brian; Brightwell, Ronald Brian

2011-09-01

Next-generation exascale systems, those capable of performing a quintillion (10{sup 18}) operations per second, are expected to be delivered in the next 8-10 years. These systems, which will be 1,000 times faster than current systems, will be of unprecedented scale. As these systems continue to grow in size, faults will become increasingly common, even over the course of small calculations. Therefore, issues such as fault tolerance and reliability will limit application scalability. Current techniques to ensure progress across faults like checkpoint/restart, the dominant fault tolerance mechanism for the last 25 years, are increasingly problematic at the scales of future systems due to their excessive overheads. In this work, we evaluate a number of techniques to decrease the overhead of checkpoint/restart and keep this method viable for future exascale systems. More specifically, this work evaluates state-machine replication to dramatically increase the checkpoint interval (the time between successive checkpoint) and hash-based, probabilistic incremental checkpointing using graphics processing units to decrease the checkpoint commit time (the time to save one checkpoint). Using a combination of empirical analysis, modeling, and simulation, we study the costs and benefits of these approaches on a wide range of parameters. These results, which cover of number of high-performance computing capability workloads, different failure distributions, hardware mean time to failures, and I/O bandwidths, show the potential benefits of these techniques for meeting the reliability demands of future exascale platforms.

Investigation of the Development of 7th Grade Students’ Skills to Define, Construct and Classify Polygons with Cabri Geometry

Directory of Open Access Journals (Sweden)

Ahmet Yanık

2013-03-01

Full Text Available The aim of the study is to investigate the development of 7th Grade students’ skills to define, construct and classify polygons in geometry course with Cabri Geometry II Plus software geometry, an example of dynamic geometry software. The study used qualitative and quantitative research methods in accordance with the research objectives and focus, so it was designed as a mixed method research. The participants of the study were 21 7th Grade students, 11 girls and 10 boys, who were attending a secondary school in Eskişehir city center during 2012-2013 school year. As a source of qualitative data, four students in this class were selected for the interview. The data were collected with “Polygon Identification and Classification Scale”, one group pre-test and post-test in order to determine the level of development and significance level of the gender variable, and Cabri Geometry worksheets developed by the researchers. The quantitative data were analyzed with SPSS Statistics 20. Also, t-test and Wilcoxon test were used in data analysis. The data obtained from the interviews were analyzed through descriptive analysis. The qualitative data showed that the mean of correct answers given by the students to the questions in the Polygon Identification and Classification Scale was higher in the post-test than the pre-test. The ttest results for the pre-test and post-test mean scores and the results of the paired samples test showed a significant difference in favor of the post-test. There was no significant difference based on the gender variable. On the other hand, the data obtained from the interviews were coded under five different themes. The activities about the concept of formation showed that incorrect formations caused incorrect generalizations about the shapes. The study found that, as a result of the teaching practice in the study, hierarchical relations among polygons were expressed correctly. Finally, after the practice, the

Advanced RESTART method for the estimation of the probability of failure of highly reliable hybrid dynamic systems

International Nuclear Information System (INIS)

Turati, Pietro; Pedroni, Nicola; Zio, Enrico

2016-01-01

The efficient estimation of system reliability characteristics is of paramount importance for many engineering applications. Real world system reliability modeling calls for the capability of treating systems that are: i) dynamic, ii) complex, iii) hybrid and iv) highly reliable. Advanced Monte Carlo (MC) methods offer a way to solve these types of problems, which are feasible according to the potentially high computational costs. In this paper, the REpetitive Simulation Trials After Reaching Thresholds (RESTART) method is employed, extending it to hybrid systems for the first time (to the authors’ knowledge). The estimation accuracy and precision of RESTART highly depend on the choice of the Importance Function (IF) indicating how close the system is to failure: in this respect, proper IFs are here originally proposed to improve the performance of RESTART for the analysis of hybrid systems. The resulting overall simulation approach is applied to estimate the probability of failure of the control system of a liquid hold-up tank and of a pump-valve subsystem subject to degradation induced by fatigue. The results are compared to those obtained by standard MC simulation and by RESTART with classical IFs available in the literature. The comparison shows the improvement in the performance obtained by our approach. - Highlights: • We consider the issue of estimating small failure probabilities in dynamic systems. • We employ the RESTART method to estimate the failure probabilities. • New Importance Functions (IFs) are introduced to increase the method performance. • We adopt two dynamic, hybrid, highly reliable systems as case studies. • A comparison with literature IFs proves the effectiveness of the new IFs.

Flowsheet for shear/leach processing of N Reactor fuel at PUREX

International Nuclear Information System (INIS)

Enghusen, M.B.

1995-01-01

This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Raickovic, N.; Radivojevic, J.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1983-12-01

After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW [sr

Status of Dalat research reactor and progress of new reactor plan in Vietnam

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Vien, Luong Ba

2005-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

RDEL: Restart Differential Evolution algorithm with Local Search Mutation for global numerical optimization

Directory of Open Access Journals (Sweden)

Ali Wagdy Mohamed

2014-11-01

Full Text Available In this paper, a novel version of Differential Evolution (DE algorithm based on a couple of local search mutation and a restart mechanism for solving global numerical optimization problems over continuous space is presented. The proposed algorithm is named as Restart Differential Evolution algorithm with Local Search Mutation (RDEL. In RDEL, inspired by Particle Swarm Optimization (PSO, a novel local mutation rule based on the position of the best and the worst individuals among the entire population of a particular generation is introduced. The novel local mutation scheme is joined with the basic mutation rule through a linear decreasing function. The proposed local mutation scheme is proven to enhance local search tendency of the basic DE and speed up the convergence. Furthermore, a restart mechanism based on random mutation scheme and a modified Breeder Genetic Algorithm (BGA mutation scheme is combined to avoid stagnation and/or premature convergence. Additionally, an exponent increased crossover probability rule and a uniform scaling factors of DE are introduced to promote the diversity of the population and to improve the search process, respectively. The performance of RDEL is investigated and compared with basic differential evolution, and state-of-the-art parameter adaptive differential evolution variants. It is discovered that the proposed modifications significantly improve the performance of DE in terms of quality of solution, efficiency and robustness.

Solution structure of the N-terminal domain of a replication restart primosome factor, PriC, in Escherichia coli

Science.gov (United States)

Aramaki, Takahiko; Abe, Yoshito; Katayama, Tsutomu; Ueda, Tadashi

2013-01-01

In eubacterial organisms, the oriC-independent primosome plays an essential role in replication restart after the dissociation of the replication DNA-protein complex by DNA damage. PriC is a key protein component in the replication restart primosome. Our recent study suggested that PriC is divided into two domains: an N-terminal and a C-terminal domain. In the present study, we determined the solution structure of the N-terminal domain, whose structure and function have remained unknown until now. The revealed structure was composed of three helices and one extended loop. We also observed chemical shift changes in the heteronuclear NMR spectrum and oligomerization in the presence of ssDNA. These abilities may contribute to the PriC-ssDNA complex, which is important for the replication restart primosome. PMID:23868391

RA reactor operation and maintenance in 1996, Part 1

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

1996-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

Fiscal year 1999 multi-year work plan, advanced reactors transition program

International Nuclear Information System (INIS)

Gantt, D.A.

1998-01-01

The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart

Fiscal year 1999 multi-year work plan, advanced reactors transition program

Energy Technology Data Exchange (ETDEWEB)

Gantt, D.A.

1998-09-17

The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart.

Multi keno-VAX a modified version of the reactor computer code Multi keno-2

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig.

Multi keno-VAX a modified version of the reactor computer code Multi keno-2

International Nuclear Information System (INIS)

Imam, M.

1995-01-01

The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig

Applications of implicit restarting in optimization and control Dan Sorensen

Energy Technology Data Exchange (ETDEWEB)

Sorensen, D. [Rice Univ., Houston, TX (United States)

1996-12-31

Implicit restarting is a technique for combining the implicitly shifted QR mechanism with a k-step Arnoldi or Lanczos factorization to obtain a truncated form of the implicitly shifted QR-iteration suitable for large scale eigenvalue problems. The software package ARPACK based upon this technique has been successfully used to solve large scale symmetric and nonsymmetric (generalized) eigenvalue problems arising from a variety of applications.

An asynchronous writing method for restart files in the gysela code in prevision of exascale systems*

Directory of Open Access Journals (Sweden)

Thomine O.

2013-12-01

Full Text Available The present work deals with an optimization procedure developed in the full-f global GYrokinetic SEmi-LAgrangian code (GYSELA. Optimizing the writing of the restart files is necessary to reduce the computing impact of crashes. These files require a very large memory space, and particularly so for very large mesh sizes. The limited bandwidth of the data pipe between the comput- ing nodes and the storage system induces a non-scalable part in the GYSELA code, which increases with the mesh size. Indeed the transfer time of RAM to data depends linearly on the files size. The necessity of non synchronized writing-in-file procedure is therefore crucial. A new GYSELA module has been developed. This asynchronous procedure allows the frequent writ- ing of the restart files, whilst preventing a severe slowing down due to the limited writing bandwidth. This method has been improved to generate a checksum control of the restart files, and automatically rerun the code in case of a crash for any cause.

Sophistication of burnup analysis system for fast reactor (2)

International Nuclear Information System (INIS)

Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

2010-10-01

Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

Activity report of Reactor Physics Division - 1989

International Nuclear Information System (INIS)

1990-01-01

The highlights of the various studies carried out during the year 1989 in Reactor Physics Division are presented in this report in the form of summaries. The topics are organised under the following subjects: (1) nuclear data evaluation, processing and validation, (2) core physics and analysis, (3) reacto r kinetics and safety analysis, (4) noise analysis, and radiation transport and shielding. It is observed that with the restart and operation of FBTR at low power for some time, some of the low power physics experiments were completed and plans and procedures for the remaining physics experiments at intermediate and high power (upto 10 MWt) have been prepared. The lists of publications by the members of Division and the Reactor Physics Seminars held during the year 19 89, are included at the end of the report. (author). refs., figs., tabs

Fast reactor development programme in France

Energy Technology Data Exchange (ETDEWEB)

Le Rigoleur, C [Direction des Reacteurs Nucleaires, CEA Centre d` Etudes de Cadarache, Saint-Paul-lez-Durance (France)

1998-04-01

First the general situation regarding production of electricity in France is briefly described. Then in the field of Fast Reactors, the main events of 1996 are presented. At the end of February 1996, the PHENIX reactor was ready for operation. After review meetings, the Safety Authority has requested safety improvements and technical demonstrations, before it examines the possibility of authorizing a new start-up of PHENIX. The year 1996 was devoted to this work. In 1996, SUPERPHENIX was characterized by excellent operation throughout the year. The reactor was restarted at the end of 1995 after a number of minor incidents. The reactor power was increased by successive steps: 30% Pn up to February 6, followed by 50% Pn up to May then 60% up to October and 90% Pn during the last months. A programmed shutdown period occurred during May, June and mid-July 1996. The reactor has been shutdown at the end of 1996 for the decenial control of the steam generators. The status of the CAPRA project, aimed at demonstrating the feasibility of a fast reactor to burn plutonium at as high a rate as possible and the status of the European Fast Reactor are presented as well as their evolution. Finally the R and D in support of the operation of PHENIX and SUPERPHENIX, in support of the ````knowledge-acquisition```` programme, and CAPRA and EFR programmes is presented, as well as the present status of the stage 2 dismantling of the RAPSODIE experimental fast reactor. (author). 4 refs, figs, 2 tabs.

Seismic, high wind, tornado, and probabilistic risk assessment of the high flux isotope reactor

International Nuclear Information System (INIS)

Harris, S.P.; Hashimoto, P.S.; Dizon, J.O.; Hashimoto, P.S.

1989-01-01

Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR). Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed

Asymptotic optimality of RESTART estimators in highly dependable systems

International Nuclear Information System (INIS)

Villén-Altamirano, J.

2014-01-01

We consider a wide class of models that includes the highly reliable Markovian systems (HRMS) often used to represent the evolution of multi-component systems in reliability settings. Repair times and component lifetimes are random variables that follow a general distribution, and the repair service adopts a priority repair rule based on system failure risk. Since crude simulation has proved to be inefficient for highly-dependable systems, the RESTART method is used for the estimation of steady-state unavailability and other reliability measures. In this method, a number of simulation retrials are performed when the process enters regions of the state space where the chance of occurrence of a rare event (e.g., a system failure) is higher. The main difficulty involved in applying this method is finding a suitable function, called the importance function, to define the regions. In this paper we introduce an importance function which, for unbalanced systems, represents a great improvement over the importance function used in previous papers. We also demonstrate the asymptotic optimality of RESTART estimators in these models. Several examples are presented to show the effectiveness of the new approach, and probabilities up to the order of 10 −42 are accurately estimated with little computational effort. - Highlights: • Rare event probabilities of highly reliable systems are estimated by simulation. • The asymptotic optimality of the application is proved. • A better importance function for highly reliable systems is provided in the paper

Clean energy for a new generation. Steam generator life cycle management and Bruce restart

International Nuclear Information System (INIS)

Newman, G.W.

2009-01-01

In the mid to late 1990s, Ontario Hydro decided to lay-up and write-down the Bruce A Nuclear Reactors. Upon transition to Bruce Power L.P., Canada's first and only private nuclear operator, new life and prospects were injected into the site, local economy and the provincial energy portfolio. The first step in this provincial power recovery initiative involved restart of Bruce Units 3 and 4 in the 2003/04 time-frame. Units 3 and 4 have performed beyond expectation during the last five-year operating interval. A combination of steam generator and fuel channel issues precluded a similar restart of Units 1 and 2. Enter the refurbishment of Bruce Units 1 and 2. This first-of-a-kind undertaking within the Canadian nuclear power industry is testament to the demonstrated industry leadership by Bruce Power L.P., their investors and the significant vendor community contribution that is supporting this major power infrastructure enhancement. Initiated as a 'turn-key' project solution separated from the operating units, this major refurbishment project has evolved to a fully managed in-house refurbishment project with the continued support from the broader vendor community. As part of this first-of-kind undertaking, Bruce Power L.P. is in the process of accomplishing such initiatives as a complete fuel channel re-tube (i.e. full core calandria and pressure tube replacement), replacement of all boilers (i.e. 16 in total) and the majority of feeder pipe replacement. Complimentary major upgrades and replacement of the remainder of plant equipment including both nuclear and non-nuclear valves, heat exchangers, electrical infrastructure, service water systems and components, all while meeting a parallel evolving/maturing regulatory environment related to achieving compliance with IAEA derived modern codes and standards. Returning to ground level, boiler replacement is a key part of the refurbishment undertaking and this further reflected a meeting of the 'old' and the 'new'. Pre

Removal of floating organic in Hanford Waste Tank 241-C-103 restart plan

International Nuclear Information System (INIS)

Wilson, T.R.; Hanson, C.

1994-01-01

The decision whether or not to remove the organic layer from Waste Tank 241-C-103 was deferred until May, 1995. The following restart plan was prepared for removal of the organic if the decision is to remove the organic from the waste tank 241-C-103

Removal of floating organic in Hanford Waste Tank 241-C-103 restart plan

Energy Technology Data Exchange (ETDEWEB)

Wilson, T.R.; Hanson, C.

1994-10-03

The decision whether or not to remove the organic layer from Waste Tank 241-C-103 was deferred until May, 1995. The following restart plan was prepared for removal of the organic if the decision is to remove the organic from the waste tank 241-C-103.

Preparing for the re-start

CERN Multimedia

2009-01-01

The end of a Council week is a good opportunity to bring you up to date with the status of the LHC, and I’m pleased to say that we had a good deal of positive news to report to the delegations today. The bottom line is that we remain on course to restart the LHC safely this year, albeit currently about 2-3 weeks later than we’d hoped at Chamonix. This Council week has seen many important developments for our future. I am particularly pleased that Council approved the Medium Term Plan and budget for 2010 as presented by the management. This is a strong vote of confidence in all of you. The President of Council is reporting on Council business in this issue of the Bulletin, so I will focus on the status of the LHC. A tremendous amount of work has been done to understand fully the splices in the LHC’s superconducting cable and copper stabilizers. One of these splices was the root cause of the incident last September that brought ...

LHC Report: A tough restart

CERN Multimedia

Jan Uythoven for the LHC team

2012-01-01

The third LHC Technical Stop of five days took place in the week of September 17. Getting back to normal operation after a technical stop  can sometimes be difficult, with debugging, testing and requalification required on the systems that have seen interventions. Folding in a selection of other problems can make for a frustrating time.   The new injector magnet is transported to the LHC. Photo: TE/ABT group. The restart experienced over the last days was one of the tougher ones. Many problems occurred, both small and large, one after the other; in the end it took until Sunday afternoon, 9 days after the end of the technical stop, to have a physics fill in the machine that delivered an initial luminosity similar to those before the technical stop. Most problems encountered were, in fact, not related to the technical stop. The technical stop consisted of the usual maintenance and consolidation of the various systems, but two items stand out: the replacement of the mirrors an...

Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

Energy Technology Data Exchange (ETDEWEB)

Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2011-05-27

The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

Restart oversight assessment of Hanford 242-A evaporator: Summary report

International Nuclear Information System (INIS)

1994-08-01

This report summarizes a January 17--28, 1994, oversight assessment of restart activities for the 242-A Evaporator at the US Department of Energy's (DOE's) Hanford Site about 25 miles northeast of Hanford, Washington. The assessment was conducted by qualified staff and consultants from the DOE Office of Environment, Safety and Health (EH). Its focus was the readiness of the facility for the resumption of safe operations, in particular those operations involved in the treatment and disposal of condensate from the evaporation of liquid radioactive waste, a key element of the tank waste remediation project administered by the DOE Richland Operations Office (DOE-RL). Overall, the assessment yielded eight programmatic concerns, supported by 38 individual findings. Of the concerns, four have already been closed, and the other four have been resolved. Results pointed up strengths in management and engineering design, as well as effective support of facility training programs by the management and operating contractor, Westinghouse Hanford Company (WHC). Weaknesses were evident, however, in conduct of operations, maintenance, and radiological practices. Furthermore, problems in the submittal and approval of Compliance Schedule Approvals--that is, WHC documentation of the status of compliance with DOE orders--were indicative of a programmatic breakdown in the DOE Order compliance process. According to the results of this assessment, there are no safety and health issues that would preclude or delay restart of the evaporator

Seismic, high wind, tornado, and probabilistic risk assessments of the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Harris, S.P.; Stover, R.L.; Hashimoto, P.S.; Dizon, J.O.

1989-01-01

Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR) Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed. 5 figs

Safety Evaluation Report related to the restart of Davis-Besse Nuclear Power Station, Unit 1, following the event of June 9, 1985 (Docket No. 50-346)

International Nuclear Information System (INIS)

1986-06-01

On June 9, 1985, the Davis-Besse Nuclear Power Station, operated by the Toledo Edison Company, experienced a partial loss of main feedwater while the plant was at 90% power. The ensuing reactor trip was followed by spurious isolation of the steam geneators which initiated a chain of events involving a number of equipment malfunctions and several operator errors ultimately interrupting all feedwater for a short period of time. By the time operators were able to restore feedwater, both steam generators had dried out. A letter from the Director of the Office of Nuclear Reactor Regulation, pursuant to 10 CFR 50.54(f) of the Commission's regulations, confirmed that the Davis-Besse facility would not be restarted without NRC approval. The letter also requested that Toledo Edison submit its program for resolving numerous concerns identified by the staff. In response, the license submitted the Davis-Besse Course of Action report. The staff has reviewed that document and other supporting material submitted by the licensee; the staff's evaluation of that information is presented in this report

The Acoustic Emission signal acquired by the microphones placed in the CABRI test device along the fourteen last R.I.A. experiments: an example of reproducible research in nuclear science

Energy Technology Data Exchange (ETDEWEB)

Laurent Pantera, Oumar Traore [CEA, DEN, DER/SRES, Cadarache, F-13108 Saint Paul lez Durance (France)

2015-07-01

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission. It is located at the Cadarache Research Centre in southern France and it is designed to act as a support to the French nuclear infrastructure. The purpose of the new testing programme termed, 'CABRI International Programme' (CIP) is to study the behaviour of PWR-type fuel rods at high burnup, equipped with an 'advanced' cladding, under Reactivity Initiated Accident (RIA) conditions (such as the scenario of a control rod ejection). Within the framework of this programme, piloted and funded by the French Institute of Nuclear Radioprotection and Safety (IRSN), ten tests are to be conducted with a frequency of two tests per year. The LPRE laboratory of the CEA which is in charge of the Preparation, realisation and breakdown of the test results studies the possibility to set up a new test analysis based on the processing of signals coming from sensors placed within the test equipment. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream of the test device. Studies showed the interest to realize temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena as the rod failure or the test shutdown (i.e. scram). The work presented in this article results from the will to consolidate these studies. Since the main phenomenon to be tracked is the fuel rod failure, the aim would be to highlight specific events which would have been precursors of the rod failure in order to use in the future these signals for further interpretation. Such an antecedent works resumption leads to a better understanding of the experimental needs and constitutes a good initial state to prepare the new very fast digital data acquisition systems. Although all the raw data are accessible in the form of text files

The Acoustic Emission signal acquired by the microphones placed in the CABRI test device along the fourteen last R.I.A. experiments: an example of reproducible research in nuclear science

International Nuclear Information System (INIS)

Laurent Pantera, Oumar Traore

2015-01-01

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission. It is located at the Cadarache Research Centre in southern France and it is designed to act as a support to the French nuclear infrastructure. The purpose of the new testing programme termed, 'CABRI International Programme' (CIP) is to study the behaviour of PWR-type fuel rods at high burnup, equipped with an 'advanced' cladding, under Reactivity Initiated Accident (RIA) conditions (such as the scenario of a control rod ejection). Within the framework of this programme, piloted and funded by the French Institute of Nuclear Radioprotection and Safety (IRSN), ten tests are to be conducted with a frequency of two tests per year. The LPRE laboratory of the CEA which is in charge of the Preparation, realisation and breakdown of the test results studies the possibility to set up a new test analysis based on the processing of signals coming from sensors placed within the test equipment. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream of the test device. Studies showed the interest to realize temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena as the rod failure or the test shutdown (i.e. scram). The work presented in this article results from the will to consolidate these studies. Since the main phenomenon to be tracked is the fuel rod failure, the aim would be to highlight specific events which would have been precursors of the rod failure in order to use in the future these signals for further interpretation. Such an antecedent works resumption leads to a better understanding of the experimental needs and constitutes a good initial state to prepare the new very fast digital data acquisition systems. Although all the raw data are accessible in the form of text files

Replication restart in UV-irradiated Escherichia coli involving pols II, III, V, PriA, RecA and RecFOR proteins.

Science.gov (United States)

Rangarajan, Savithri; Woodgate, Roger; Goodman, Myron F

2002-02-01

In Escherichia coli, UV-irradiated cells resume DNA synthesis after a transient inhibition by a process called replication restart. To elucidate the role of several key proteins involved in this process, we have analysed the time dependence of replication restart in strains carrying a combination of mutations in lexA, recA, polB (pol II), umuDC (pol V), priA, dnaC, recF, recO or recR. We find that both pol II and the origin-independent primosome-assembling function of PriA are essential for the immediate recovery of DNA synthesis after UV irradiation. In their absence, translesion replication or 'replication readthrough' occurs approximately 50 min after UV and is pol V-dependent. In a wild-type, lexA+ background, mutations in recF, recO or recR block both pathways. Similar results were obtained with a lexA(Def) recF strain. However, lexA(Def) recO or lexA(Def) recR strains, although unable to facilitate PriA-pol II-dependent restart, were able to perform pol V-dependent readthrough. The defects in restart attributed to mutations in recF, recO or recR were suppressed in a recA730 lexA(Def) strain expressing constitutively activated RecA (RecA*). Our data suggest that in a wild-type background, RecF, O and R are important for the induction of the SOS response and the formation of RecA*-dependent recombination intermediates necessary for PriA/Pol II-dependent replication restart. In con-trast, only RecF is required for the activation of RecA that leads to the formation of pol V (UmuD'2C) and facilitates replication readthrough.

ALICE & LHCb: refinements for the restart

CERN Multimedia

2009-01-01

Following the previous issue, the Bulletin continues its series to find out what the six LHC experiments have been up to since last September, and how they are preparing for the restart. Previously we looked at CMS and ATLAS; this issue we will round up the past 10 months of activity at ALICE and LHCb. LHCb The cavern of the LHCb experiment. This year has given LHCb the chance to install the 5th and final plane of muon chambers, which will improve the triggering at nominal luminosity. This is the final piece of the experiment to be installed. "Now the detector looks exactly as it does in the technical design report," confirms Andrei Golutvin, LHCb Spokesperson. "We also took advantage of this shutdown to make several improvements. For example, we modified the high voltage system of the electromagnetic calorimeter to reduce noise further to a negligible level. We also took some measures to improve ...

78 FR 69367 - Golden Valley Electric Association: Healy Power Plant Unit #2 Restart

Science.gov (United States)

2013-11-19

... of Decision. SUMMARY: The Rural Utilities Service (RUS) has issued a Record of Decision (ROD) for the... financing from RUS to facilitate the restart of Unit 2 and for improvements to the Healy Plant, which... DOE and AIDEA. The decision documented in RUS's ROD is that RUS agrees to consider, subject to...

The design and implementation of Berkeley Lab's linuxcheckpoint/restart

Energy Technology Data Exchange (ETDEWEB)

Duell, Jason

2005-04-30

This paper describes Berkeley Linux Checkpoint/Restart (BLCR), a linux kernel module that allows system-level checkpoints on a variety of Linux systems. BLCR can be used either as a stand alone system for checkpointing applications on a single machine, or as a component by a scheduling system or parallel communication library for checkpointing and restoring parallel jobs running on multiple machines. Integration with Message Passing Interface (MPI) and other parallel systems is described.

Fast reactor development programme in France during 1995

International Nuclear Information System (INIS)

Le Rigoleur, C.

1996-01-01

In 1995, the total amount of electricity produced in France was 471 TWh, out of which 358.2 TWh (76 %) were produced by nuclear power plants, 36.9 TWh (7.8 %) by conventional thermal plants, and 75.5 TWh (16 %) by hydraulic plants. The net electrical power consumption was 368.7 TWh. At the end of 1995, 'Electricite de France' had 54 PWR units in operation. The availability factor for these units was maintained at 81%. 1995 was marked by a decrease of unexpected shutdowns (1.8% in 1995 instead of 2.2% in 1994), a new reduction in programmed shutdown periods, and a good safety level was maintained. In the field of Fast Reactors, the main events of 1995 were the following. At the end of December 1994, the PHENIX reactor was authorized to perform its 49th cycle at 350 MW th (143 MWe). This 49th cycle was completed without any significant problems on April 7, 1995. During the remainder of the year, the reactor had been shut down in order to carry out several tasks within the scope of the ten-year extension of the PHENIX reactor's lifetime. Concerning the CREYS-MALVILLE plant (SUPER-PHENIX) the first part of the year was devoted to repairing argon leak of one of the IHX. Authorization to restart the reactor was given on August 22. The end of the year was beset by a number of minor incidents. The reactor was restarted at the end of 1995 and reactor power was increased by successive steps (30% Pn (Nominal Power) up to February 6 1996; followed by 50 %...). The 'Decret d'Autorisation de Creation' stipulates that because of its prototype character, SUPER PHENIX will have to be operated under conditions explicitly giving priority to safety and knowledge acquisition, with an objective of research and demonstration. In this context, the so-called 'knowledge acquisition' programme designed to prove the capacity of a large FBR to produce electricity on an industrial scale, to test the consumption of plutonium and minor actinides in a large fast reactor, as well as to provide

Influence of initial conditions on rod behaviour during boiling crisis phase following a reactivity initiated accident

International Nuclear Information System (INIS)

Georgenthum, V.; Sugiyama, T.

2010-01-01

In the frame of their research programs on high burn-up fuel safety, the French Institute for Radioprotection and Nuclear Safety (IRSN) and the Japan Atomic Energy Agency (JAEA) performed a large set of tests devoted to the study of PWR fuel rod behavior during Reactivity Initiated Accident (RIA) respectively in the CABRI reactor and in the NSRR reactor. The reactor test conditions are different in terms of coolant nature, temperature and pressure. In the CABRI reactor, tests were performed until now with sodium coolant at 280 Celsius degrees and 3 bar. In the NSRR reactor most of the tests were performed with stagnant water at 20 C. degrees and atmospheric pressure but recently a new high temperature high pressure capsule has been developed which allows to performed tests at up to 280 Celsius degrees and 70 bar. The paper discusses the influence of test conditions on rod behaviour during boiling phase, based on tests results and SCANAIR code calculations. The study shows that when the boiling crisis is reached, the initial inner and outer rod pressure have an essential impact on the clad straining and possible ballooning. The analysis of the different test conditions makes it possible to discriminate the influence of initial conditions on the different phases of the transient and is useful for modelling and code development. (authors)

Economic evaluation of strategies for restarting anticoagulation therapy after a first event of unprovoked venous thromboembolism.

Science.gov (United States)

Monahan, M; Ensor, J; Moore, D; Fitzmaurice, D; Jowett, S

2017-08-01

Essentials Correct duration of treatment after a first unprovoked venous thromboembolism (VTE) is unknown. We assessed when restarting anticoagulation was worthwhile based on patient risk of recurrent VTE. When the risk over a one-year period is 17.5%, restarting is cost-effective. However, sensitivity analyses indicate large uncertainty in the estimates. Background Following at least 3 months of anticoagulation therapy after a first unprovoked venous thromboembolism (VTE), there is uncertainty about the duration of therapy. Further anticoagulation therapy reduces the risk of having a potentially fatal recurrent VTE but at the expense of a higher risk of bleeding, which can also be fatal. Objective An economic evaluation sought to estimate the long-term cost-effectiveness of using a decision rule for restarting anticoagulation therapy vs. no extension of therapy in patients based on their risk of a further unprovoked VTE. Methods A Markov patient-level simulation model was developed, which adopted a lifetime time horizon with monthly time cycles and was from a UK National Health Service (NHS)/Personal Social Services (PSS) perspective. Results Base-case model results suggest that treating patients with a predicted 1 year VTE risk of 17.5% or higher may be cost-effective if decision makers are willing to pay up to £20 000 per quality adjusted life year (QALY) gained. However, probabilistic sensitivity analysis shows that the model was highly sensitive to overall parameter uncertainty and caution is warranted in selecting the optimal decision rule on cost-effectiveness grounds. Univariate sensitivity analyses indicate variables such as anticoagulation therapy disutility and mortality risks were very influential in driving model results. Conclusion This represents the first economic model to consider the use of a decision rule for restarting therapy for unprovoked VTE patients. Better data are required to predict long-term bleeding risks during therapy in this

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

Science.gov (United States)

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

The RA nuclear research reactor at VINCA Institute as an engineering and scientific challenge

International Nuclear Information System (INIS)

Mesarovic, M.

1997-01-01

The RA nuclear research at the Vinca Institute of Nuclear Sciences is the largest nuclear research facility in Yugoslavia and belongs to that generation of research reactors which have had an important contribution to nuclear technology development. As these older reactors were generally not built to specific nuclear standards, new safety systems had to be installed at the RA reactor for a renewal of its operating licence in 1984 and it was shut down, after 25 years of operation. Although all the required and several additional systems were built for the restart of the RA reactor, a disruption of foreign delivery of new control equipment caused its conversion to a 'dormant' facility, and it is still out of operation. Therefore, the future status of the RA reactor presents an engineering and scientific challenge to the engineers and scientists from Yugoslavia and other countries that may be interested to participate. To attract their attention on the subject, principal features of the RA reactor and its present status are described in detail, based on a recent engineering economic and safety evaluation. A comparative review of the world research reactors is also presented.(author)

Validation of DRAGON side-step method for Bruce-A restart Phase-B physics tests

International Nuclear Information System (INIS)

Shen, W.; Ngo-Trong, C.; Davis, R.S.

2004-01-01

The DRAGON side-step method, developed at AECL, has a number of advantages over the all-DRAGON method that was used before. It is now the qualified method for reactivity-device calculations. Although the side-step-method-generated incremental cross sections have been validated against those previously calculated with the all-DRAGON method, it is highly desirable to validate the side-step method against device-worth measurements in power reactors directly. In this paper, the DRAGON side-step method was validated by comparison with the device-calibration measurements made in Bruce-A NGS Unit 4 restart Phase-B commissioning in 2003. The validation exercise showed excellent results, with the DRAGON code overestimating the measured ZCR worth by ∼5%. A sensitivity study was also performed in this paper to assess the effect of various DRAGON modelling techniques on the incremental cross sections. The assessment shows that the refinement of meshes in 3-D and the use of the side-step method are two major reasons contributing to the improved agreement between the calculated ZCR worths and the measurements. Use of different DRAGON versions, DRAGON libraries, local-parameter core conditions, and weighting techniques for the homogenization of tube clusters inside the ZCR have a very small effect on the ZCR incremental thermal absorption cross section and ZCR reactivity worth. (author)

TMI-1 restart: an evaluation of the licensee's management integrity as it affects restart of Three Mile Island Nuclear Station (Unit 1 Docket 50-289). Supplement 5

International Nuclear Information System (INIS)

1984-07-01

Supplement 5 to the Safety Evaluation Report (SER) on TMI-1 Restart documents the review by the Nuclear Regulatory Commission (NRC) staff of nine investigations conducted by the NRC Office of Investigations into matters identified as relevant and material to an evaluation of the licensee's management integrity. The staff has included, as part of its evaluation, materials from its review of the GPU v. B and W lawsuit record (NUREG-1020LD, GPU, v. B and W Lawsuit Review and Its Effect on TMI-1) as well as other relevant materials developed since the close of the record in the TMI-1 Restart proceeding. In developing its position on General Public Utilities Nuclear Corporation's character (i.e., management integrity), the staff evaluated matters that cast doubt on the licensee's character, individually and collectively; considered the remedial actions taken by the licensee; and balanced past improper conduct of the licensee against its subsequent record of remedial actions and performance and record of current senior management of the licensee. The staff concluded that, while the past improper conduct was grave, the remedial actions taken, the subsequent record of performance, and the record of current senior management support a finding that GPUN can and will operate TMI-1 without undue risk to the health and safety of the public

Rapid restoration of methanogenesis in an acidified UASB reactor treating 2,4,6-trichlorophenol (TCP).

Science.gov (United States)

Díaz-Báez, María Consuelo; Valderrama-Rincon, Juan Daniel

2017-02-15

Anaerobic bioreactors are often used for removal of xenobiotic and highly toxic pollutants from wastewater. Most of the time, the pollutant is so toxic that the stability of the reactor becomes compromised. It is well known that methanogens are one of the most sensitive organisms in the anaerobic consortia and hence the stability of the reactors is highly dependant on methanogenesis. Unfortunately few studies have focused on recovering the methanogenic activity once it has been inhibited by highly toxic pollutants. Here we establish a quick recovery strategy for neutralization of an acidified UASB reactor after failure by intoxication with an excess of TCP in the influent. Once the reactor returned to pH values compatible with methanogenesis, biogas production was re-started after one day and the system was re-acclimated to TCP. Successful removal of TCP from synthetic wastewater was shown for concentrations up to 70mg/L after restoration. Copyright © 2016 Elsevier B.V. All rights reserved.

Press Conference: LHC Restart, Season 2

CERN Multimedia

CERN. Geneva

2015-01-01

PRESS BRIEFING ON THE LARGE HADRON COLLIDER (LHC) RE-START, SEASON 2 AT CERN, GLOBE OF SCIENCE AND INNOVATION Where :   http://cern.ch/directions   at the Globe of Science and Innovation When : Thursday, 12 March from 2.30 to 3.30pm - Open seating as from 2.15pm Speakers : CERN’s Director General, Rolf Heuer and Director of Accelerators, Frédérick Bordry, and representatives of the LHC experiments Webcast : https://webcast.web.cern.ch/webcast/ Dear Journalists, CERN is pleased to invite you to the above press briefing which will take place on Thursday 12 March, in the Globe of Science and Innovation, 1st floor, from 2.30 to 3.30pm. The Large Hadron Collider (LHC) is ready to start up for its second three-year run. The 27km LHC is the largest and most powerful particle accelerator in the world operating at a temperature of -217 degrees Centigrade and powered to a current of 11,000 amps. Run 2 of the LHC follows a two-year technical s...

Promoting Recruitment using Information Management Efficiently (PRIME): a stepped-wedge, cluster randomised trial of a complex recruitment intervention embedded within the REstart or Stop Antithrombotics Randomised Trial.

Science.gov (United States)

Maxwell, Amy E; Parker, Richard A; Drever, Jonathan; Rudd, Anthony; Dennis, Martin S; Weir, Christopher J; Al-Shahi Salman, Rustam

2017-12-28

Few interventions are proven to increase recruitment in clinical trials. Recruitment to RESTART, a randomised controlled trial of secondary prevention after stroke due to intracerebral haemorrhage, has been slower than expected. Therefore, we sought to investigate an intervention to boost recruitment to RESTART. We conducted a stepped-wedge, cluster randomised trial of a complex intervention to increase recruitment, embedded within the RESTART trial. The primary objective was to investigate if the PRIME complex intervention (a recruitment co-ordinator who conducts a recruitment review, provides access to bespoke stroke audit data exports, and conducts a follow-up review after 6 months) increases the recruitment rate to RESTART. We included 72 hospital sites located in England, Wales, or Scotland that were active in RESTART in June 2015. All sites began in the control state and were allocated using block randomisation stratified by hospital location (Scotland versus England/Wales) to start the complex intervention in one of 12 different months. The primary outcome was the number of patients randomised into RESTART per month per site. We quantified the effect of the complex intervention on the primary outcome using a negative binomial, mixed model adjusting for site, December/January months, site location, and background time trends in recruitment rate. We recruited and randomised 72 sites and recorded their monthly recruitment to RESTART over 24 months (March 2015 to February 2017 inclusive), providing 1728 site-months of observations for the primary analysis. The adjusted rate ratio for the number of patients randomised per month after allocation to the PRIME complex intervention versus control time before allocation to the PRIME complex intervention was 1.06 (95% confidence interval 0.55 to 2.03, p = 0.87). Although two thirds of respondents to the 6-month follow-up questionnaire agreed that the audit reports were useful, only six patients were reported to

Stop and Restart Effects on Modern Vehicle Starting System Components

Energy Technology Data Exchange (ETDEWEB)

Windover, Paul R. [Argonne National Lab. (ANL), Argonne, IL (United States); Owens, Russell J. [Argonne National Lab. (ANL), Argonne, IL (United States); Levinson, Terry M. [Argonne National Lab. (ANL), Argonne, IL (United States); Laughlin, Michael [Argonne National Lab. (ANL), Argonne, IL (United States); Gaines, Linda [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-01-01

Many drivers of personal and commercial vehicles believe that turning the vehicle off and on frequently instead of idling will cause premature wear of the starter system (starter motor and starter battery). As a result, they are concerned that the replacement cost of the starter motor and/or battery due to increased manual engine cycling would be more than the cumulative cost of the fuel saved by not idling unnecessarily. A number of variables play a role in addressing this complex concern, including the number of starting cycles per day, the time between starting cycles, the intended design life of the starting system, the amount of fuel used to restart an engine, and the cumulative cost of the saved fuel. Qualitative and quantitative information from a variety of sources was used to develop a life-cycle economic model to evaluate the cost and quantify the realistic factors that are related to the permissible frequency of starter motor cycles for the average vehicle to economically minimize engine idle time. Annual cost savings can be calculated depending on shutdown duration and the number of shutdown cycles per day. Analysis shows that cost savings are realized by eliminating idling exceeding one minute by shutting down the engine and restarting it. For a typical motorist, the damage to starting system components resulting from additional daily start cycles will be negligible. Overall, it was found that starter life is mostly dependent on the total number of start cycles, while battery life is more dependent on ensuring a full charge between start events.

Restart of the chemical preparation process for the fabrication of ZnO varistors for ferroelectric neutron generator power supplies

International Nuclear Information System (INIS)

Lockwood, Steven John

2005-01-01

To date, all varistors used in ferroelectric neutron generators have been supplied from a single, proprietary source, General Electric Corporate Research and Development (GE CR and D). To protect against the vulnerability of a single source, Sandia initiated a program in the early 1980's to develop a second source for this material. A chemical preparation process for making homogeneous, high purity ZnO-based varistor powder was generated, scaled to production quantities, and transferred to external suppliers. In 1992, the chem-prep varistor program was suspended when it appeared there was sufficient inventory of GE CR and D material to supply ferroelectric neutron generator production for many years. In 1999, neutron generator production schedules increased substantially, resulting in a predicted exhaustion of the existing supply of varistor material within five years. The chem-prep program was restarted in January, 2000. The goals of the program were to (1) duplicate the chem-prep powder synthesis process that had been qualified for WR production, (2) demonstrate sintered billets from the chem-prep powder met requirements, (3) develop a process for rod fabrication and demonstrate that all component specifications could be met, and (4) optimize the process from powder synthesis through component fabrication for full-scale production. The first three of these goals have been met and are discussed in this report. A facility for the fabrication of production quantities of chem-prep powder has been established. All batches since the restart have met compositional requirements, but differences in sintering behavior between the original process and the restarted process were noted. Investigation into the equipment, precipitant stoichiometry, and powder processing procedures were not able to resolve the discrepancies. It was determined that the restarted process, which incorporated Na doping for electrical stability (a process that was not introduced until the end of the

RESTART simulation of non-Markov consecutive-k-out-of-n: F repairable systems

International Nuclear Information System (INIS)

Villen-Altamirano, Jose

2010-01-01

The reliability of consecutive-k-out-of-n: F repairable systems and (k-1)-step Markov dependence is studied. The model analyzed in this paper is more general than those of previous studies given that repair time and component lifetimes are random variables that follow a general distribution. The system has one repair service which adopts a priority repair rule based on system failure risk. Since crude simulation has proved to be inefficient for highly dependable systems, the RESTART method was used for the estimation of steady-state unavailability, MTBF and unreliability. Probabilities up to the order of 10 -16 have been accurately estimated with little computational effort. In this method, a number of simulation retrials are performed when the process enters regions of the state space where the chance of occurrence of a rare event (e.g., a system failure) is higher. The main difficulty for the application of this method is to find a suitable function, called the importance function, to define the regions. Given the simplicity involved in changing some model assumptions with RESTART, the importance function used in this paper could be useful for dependability estimation of many systems.

On factoring RSA modulus using random-restart hill-climbing algorithm and Pollard’s rho algorithm

Science.gov (United States)

Budiman, M. A.; Rachmawati, D.

2017-12-01

The security of the widely-used RSA public key cryptography algorithm depends on the difficulty of factoring a big integer into two large prime numbers. For many years, the integer factorization problem has been intensively and extensively studied in the field of number theory. As a result, a lot of deterministic algorithms such as Euler’s algorithm, Kraitchik’s, and variants of Pollard’s algorithms have been researched comprehensively. Our study takes a rather uncommon approach: rather than making use of intensive number theories, we attempt to factorize RSA modulus n by using random-restart hill-climbing algorithm, which belongs the class of metaheuristic algorithms. The factorization time of RSA moduli with different lengths is recorded and compared with the factorization time of Pollard’s rho algorithm, which is a deterministic algorithm. Our experimental results indicates that while random-restart hill-climbing algorithm is an acceptable candidate to factorize smaller RSA moduli, the factorization speed is much slower than that of Pollard’s rho algorithm.

The pros and cons about restarting and awareness about nuclear power generation. Further findings from INSS's analysis of the opinion survey answers

International Nuclear Information System (INIS)

Kitada, Atsuko

2015-01-01

In this paper, the pros and cons and the awareness of their background on the restart of nuclear power plants (NPPs) were analyzed based on the data of opinion polls conducted by news organizations and INSS. The results were as follows: (1) Although opposition to restart has been nearly 60% in the case when the question has only 2 choices of pros and cons, the 60% includes many people who “cannot say either way”. (2) For approval of restarting, it is necessary to have the attitude to tolerate the use of nuclear power generation (NPG), and it is extremely important that people think “safety has been confirmed,” but it does not seem to be enough reason to hesitate in the restart. (3) From the open-ended question about the influences of long-term shutdown of NPPs on the respondents and Japanese society, 50% of the respondents said there was no influence. 20% said they felt safe or easy because no accident could occur. Only 20% described economic influence, such as a real increase in electricity cost, especially among men or the people who were in favor of restarting. (4) When the above-described influences, people's recognition of the utility of NPG (or thinking about the problems that occur when reducing NPG) and people's criteria for selecting an electric power generation method were organized in terms of the 3Es (Energy security, Economic efficiency, Environment), there was quite a difference from the idea of energy policy that emphasizes a good balance among these 3Es. People's recognition of the utility of NPG has been declining. This was considered to be the reason that neither the present influences nor the importance of the invisible influence at a macro-level had been recognized. For restarting to be supported by many people, there is a need for appropriate information about not only the safety, but also the importance from the viewpoints of the 3Es and the benefits of NPG in the 3Es. (author)

TOTEM and LHCf: refinements for the restart

CERN Multimedia

2009-01-01

Following the previous two issues, the Bulletin continues its series to find out what the six LHC experiments have been up to since last September, and how they are preparing for the restart. We covered CMS, ATLAS, LHCb and ALICE in previous issues. In this issue we will round up the past 10 months of activity at TOTEM and LHCf. Roman Pots of the TOTEM experiment.TOTEM The past 10 months at TOTEM have been amongst the busiest since the project’s inception. The delay in the LHC startup has certainly had a silver lining for the TOTEM collaboration - not only has it given them a much-needed opportunity to test and install many crucial new detector parts, but also the lower energy range that the LHC will initially operate at in 2009 is perfect for TOTEM physics. "In fact, the LHC almost seems to be following the schedule of TOTEM!" jokes Karsten Eggert, TOTEM spokesperson. TOTEM is made up of three different detectors spread out...

U.S. uranium supply to the research and test reactor community

International Nuclear Information System (INIS)

Parker, Elaine M.

2002-01-01

From the 1950s through the early 1990s, the U.S. Department of Energy (DOE) was the primary supplier of low enriched uranium (LEU) and highly enriched uranium (HEU) to research and test reactors worldwide. The formerly called Y-12 Plant in Oak Ridge, Tennessee, was put into operational stand down in 1994 due to inadequate safety documentation. This paper will discuss the re-start of the Y-12 Plant and its current capabilities. Additionally, the paper will address recent changes within the DOE, with the creation of the National Nuclear Security Administration (NNSA). It will show how the change to NNSA and an organizational re-alignment has improved efficiencies. NNSA is committed to operate its sales program so that it is complementary to, and in support of, the Reduced Enrichment for Research and Test Reactors (RERTR) and Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Return Programs. The NNSA is committed to provide an assurance of competitively-priced, high-quality uranium supply to the research and test reactor community under long-term contracts. This paper will discuss some of NNSA's recent successes in long-term contracting and meeting deliveries. (author)

Experimental facilities and simulation means

International Nuclear Information System (INIS)

Thomas, J.B.

2009-01-01

This paper and its associated series of slides review the experimental facilities and the simulation means used for the development of nuclear reactors in France. These experimental facilities include installations used for the measurement and qualification of nuclear data (mainly cross-sections) like EOLE reactor and Minerve zero power reactor, installations like material testing reactors, installations dedicated to reactor safety experiments like Cabri reactor, and other installations like accelerators (Jannus accelerator, GANIL for instance) that are complementary to neutron irradiations in experimental reactors. The simulation means rely on a series of advanced computer codes: Tripoli-Apollo for neutron transport, Numodis for irradiation impact on materials, Neptune and Cathare for 2-phase fluid dynamics, Europlexus for mechanical structures, and Pleiades (with Alcyone) for nuclear fuels. (A.C.)

A study on people's awareness about the restarting and decommissioning of nuclear power plants

International Nuclear Information System (INIS)

Goto, Manabu; Sakai, Yukimi

2015-01-01

In this study, we conducted two questionnaire surveys targeting a total of 918 respondents living in the cities of Kyoto, Osaka and Kobe, in order to elucidate people's awareness of three things: 1) restart of nuclear power plants; 2) extension of the operation period of aging plants; and 3) decommissioning. The results are as follows: 1) People who think that electrical power companies voluntarily take higher safety measures trust the power companies and do not oppose the restart of the nuclear power plants, as compared to people who think that power companies only meet the requirements set by the nuclear regulatory agency. 2) When people were given information about aging measures and conforming to new regulatory standards, their anxiety toward the operation of aging plants was reduced. 3) People thought that decommissioning work was important for society. However, a small number of people thought it was a job worthwhile doing. (author)

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

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

2010-01-01

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

Rheology and FTIR studies of model waxy crude oils with relevance to gelled pipeline restart

Energy Technology Data Exchange (ETDEWEB)

Magda, J.J.; Guimeraes, K.; Deo, M.D. [Utah Univ., Salt Lake City, UT (United States). Dept. of Chemical Engineering; Venkatesan, R.; Montesi, A. [Chevron Energy Technology Co., Houston, TX (United States)

2008-07-01

Gels composed of wax crystals may sometimes form when crude oils are transported in pipelines when ambient temperatures are low. The gels may stop the pipe flow, making it difficult or even impossible to restart the flow without breaking the pipe. Rheology and FTIR techniques were used to study the problem and to characterize transparent model waxy crude oils in pipeline flow experiments. These model oils were formulated without any highly volatile components to enhance the reproducibility of the rheology tests. Results were presented for the time- and temperature-dependent rheology of the model waxy crude oils as obtained in linear oscillatory shear and in creep-recovery experiments. The model oils were shown to exhibit many of the rheological features reported for real crude oils, such as 3 distinct apparent yield stresses, notably static yield stress, dynamic yield stress, and elastic-limit yield stress. It was concluded that of the 3, the static yield stress value, particularly its time dependence, can best be used to predict the restart behaviour observed for the same gel in model pipelines.

Understanding Victims of Technological Disaster: Beliefs and Worries of Three Mile Island.

Science.gov (United States)

Prince-Embury, Sandra; Rooney, James

The primary purpose of the present study was to examine how prevalent were concerns about restarting Three Mile Island nuclear reactor Unit I among people within a five-mile radius of the plant four years after the accident involving reactor Unit II. Also explored were concerns related to expectations about the restart of Unit I, perception of…

Modeling of the re-starting of waxy crude oil flows in pipelines; Modelisation du redemarrage des ecoulements de bruts paraffiniques dans les conduites petrolieres

Energy Technology Data Exchange (ETDEWEB)

Vinay, G.

2005-11-15

Pipelining crude oils that contain large proportions of paraffins can cause many specific difficulties. These oils, known as waxy crude oils, usually exhibit high 'pour point', where this temperature is higher than the external temperature conditions surrounding the pipeline. During the shutdown, since the temperature decreases in the pipeline, the gel-like structure builds up and the main difficulty concerns the issue of restarting. This PhD attempts to improve waxy crude oil behaviour understanding thanks to experiment, modelling and numerical simulation in order to predict more accurately time and pressure required to restart the flow. Using various contributions to the literature, waxy crude oils are described as viscoplastic, thixotropic and compressible fluid. Strong temperature history dependence plays a prevailing role in the whole shutdown and restart process. Thus, waxy crude oils under flowing conditions correspond to the non-isothermal flow of a viscoplastic material with temperature-dependent rheological properties. Besides, the restart of a waxy crude oil is simulated by the isothermal transient flow of a weakly compressible thixotropic fluid in axisymmetric pipe geometry. We retain the Houska model to describe the thixotropic/viscoplastic feature of the fluid and compressibility is introduced in the continuity equation. The viscoplastic constitutive equation is involved using an augmented Lagrangian method and the resulting equivalent saddle-point problem is solved thanks to an Uzawa-like algorithm. Governing equations are discretized using a Finite Volume method and the convection terms are treated thanks to a TVD (Total Variation Diminishing) scheme. The Lagrangian functional technique usually used for incompressible viscoplastic flows, is adapted to compressible situations. Several numerical results attest the good convergence properties of the proposed transient algorithm. The non-isothermal results highlight the strong sensitivity of

Restart Plan for the Prototype Vertical Denitration Calciner [SD Coversheet has Incorrect Document Number

Energy Technology Data Exchange (ETDEWEB)

SUTTER, C.S.

1999-07-26

Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing.

Management of safety and risk at the HFIR [High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Glovier, H.A.

1990-01-01

This paper discusses the management of safety and risk at the High-Flux Isotope Reactor (HFIR), a category A research reactor at Oak Ridge National Laboratory (ORNL). The HFIR went critical in 1966 and operated at its designed 100 MW for 20 yr until it was shut down on November 14, 1986. It operated at a >90% availability and without significant event during this period. The result was a complacent management program lacking rigor. This complacency came to an end with the Chernobyl accident, which led to the appointment of an internal committee to assess the safety of ORNL reactor operations. This committee found that HFIR pressure vessel material specimens removed several years earlier had not been analyzed. This issue led to a general review of management practices that were found lacking in quality assurance, safety documentation, training process, and emergency planning, among others. Management accountability was lacking, as shown by design basis and safety analyses that were not up to data and by the fact that reactor operators whose requalification examinations had not been graded were allowed to continue operating the reactor over a long period of time. Between shutdown in 1986 and restart in April 1989, significant management changes and initiatives were made in the area of risk and safety management of ORNL reactors. These are presented briefly in this paper

The accuracy evaluation according to dose delivery interruption and restart for volumetric modulated arc therapy

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Hyung; Bae, Sun Myung; Kwak, Jung Won; Kang, Tae Young; Bck, Geum Mun [Dept. of Radiation Oncology, Asan Medical Center, Seoul(Korea, Republic of)

2013-03-15

The accurate movement of gantry rotation, collimator and correct application of dose rate are very important to approach the successful performance of Volumetric Modulated Arc Therapy (VMAT), because it is tightly interlocked with a complex treatment plan. The interruption and restart of dose delivery, however, are able to occur on treatment by various factors of a treatment machine and treatment plan. If unexpected problems of a treat machine or a patient interrupt the VMAT, the movement of treatment machine for delivering the remaining dose will be restarted at the start point. In this investigation, We would like to know the effect of interruptions and restart regarding dose delivery at VMAT. Treatment plans of 10 patients who had been treated at our center were used to measure and compare the dose distribution of each VMAT after converting to a form of digital image and communications in Medicine (DICOM) with treatment planning system (Eclipse V 10.0, Varian, USA). We selected the 6 MV photon energy of Trilogy (Varian, USA) and used OmniPro I'mRT system (V 1.7b, IBA dosimetry, Germany) to analyze the data that were acquired through this measurement with two types of interruptions four times for each case. The door interlock and the beam-off were used to stop and then to restart the dose delivery of VMAT. The gamma index in OmniPro I'mRT system and T-test in Microsoft Excel 2007 were used to evaluate the result of this investigation. The deviations of average gamma index in cases with door interlock, beam-off and without interruption on VMAT are 0.141, 0.128 and 0.1. The standard deviations of acquired gamma values are 0.099, 0.091, 0.071 and The maximum gamma value in each case is 0.413, 0.379, 0.286, respectively. This analysis has a 95-percent confidence level and the P-value of T-test is under 0.05. Gamma pass rate (3%, 3 mm) is acceptable in all of measurements. As a result, We could make sure that the interruption of this investgation are not

Psychological adaptation among residents following restart of Three Mile Island.

Science.gov (United States)

Prince-Embury, S; Rooney, J F

1995-01-01

Psychological adaptation is examined in a sample of residents who remained in the vicinity of Three Mile Island following the restart of the nuclear generating facility which had been shut down since the 1979 accident. Findings indicate a lowering of psychological symptoms between 1985 and 1989 in spite of increased lack of control, less faith in experts and increased fear of developing cancer. The suggestion is made that reduced stress might have been related to a process of adaptation whereby a cognition of emergency preparedness was integrated by some of these residents as a modulating cognitive element. Findings also indicate that "loss of faith in experts" is a persistently salient cognition consistent with the "shattered assumptions" theory of victimization.

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

International Nuclear Information System (INIS)

2016-02-01

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

Simulated annealing with restart strategy for the blood pickup routing problem

Science.gov (United States)

Yu, V. F.; Iswari, T.; Normasari, N. M. E.; Asih, A. M. S.; Ting, H.

2018-04-01

This study develops a simulated annealing heuristic with restart strategy (SA_RS) for solving the blood pickup routing problem (BPRP). BPRP minimizes the total length of the routes for blood bag collection between a blood bank and a set of donation sites, each associated with a time window constraint that must be observed. The proposed SA_RS is implemented in C++ and tested on benchmark instances of the vehicle routing problem with time windows to verify its performance. The algorithm is then tested on some newly generated BPRP instances and the results are compared with those obtained by CPLEX. Experimental results show that the proposed SA_RS heuristic effectively solves BPRP.

Stress corrosion cracking studies of reactor pressure vessel steels. Final report

International Nuclear Information System (INIS)

Van Der Sluys, W.A.

1996-10-01

The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

Importance of Entrepreneurs’ Knowledge for Business Restarts of Micro and Small Enterprises

OpenAIRE

Ropęga, Jarosław

2013-01-01

Przedstawiona publikacja jest poświęcona charakterystyce i znaczeniu restartów jako następstw niepowodzeń gospodarczych. Zjawisko to staje się coraz częściej wymieniane w kategorii badawczej. Jest ona wynikiem dostrzeżenia potrzeby dyskusji nad postawami osób, które po wcześniejszych doświadczeniach zakończonych niepowodzeniem biznesu chcą ponownie rozpocząć nową działalność gospodarczą. W pierwszej części artykułu omówiono pojęcie oraz skalę tego zjawiska, wskazując na jego ...

Current status of JMTR for restart

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Kimura, Nobuaki; Ooka, Makoto

2013-01-01

After the 2011 off the Pacific Coast of Tohoku Earthquake on March 11, 2011, JMTR has been challenging to the inspection of facility equipment, seismic soundness evaluation through earthquake response analysis, repair of facilities, correspondence to the report matters stipulated by laws and regulations, and improvement of irradiation facilities, and based on these, it have formulated the operation schedule from FY2013. In the future, JMTR will explain the soundness of the facilities to the Nuclear Regulatory Commission, and receive the facility's regular inspection after the completion of piping updating work related to the report matters stipulated by laws and regulations. After obtaining the understanding of local municipalities, it aims to restart the facilities in August 2013. After the reoperation, it will make efforts to safely and stably operate the facilities with a target of the operating rate of world top class. In addition, the following challenges are planned: (1) improvement of the facilities based on external funds for expanding utilization, (2) human resource development in the nuclear field by utilizing JMTR, and (3) development of an advanced monitoring system for improving the safety of nuclear power plants by utilizing the irradiation technology that has been accumulated in JMTR. With the aim for JMTR becoming the international hub, JMTR will continue aggressive activities. (A.O.)

A proposal on restart rule of nuclear power plants with piping having local wall thinning subjected to an earthquake. Former part. Aiming at further application

International Nuclear Information System (INIS)

Urabe, Yoshio

2011-01-01

Restart rule of nuclear power plants (NPPs) with piping having local wall thinning subjected to an earthquake was proposed taking account of local wall thinning, seismic effects and restart of NPPs with applicability of 'Guidelines for NPP Response to an Earthquake (EPRI NP-6695)' in Japan. Japan Earthquake Damage Intensity Scale (JEDIS) and Earthquake Ground Motion Level (EGML) were introduced. JEDIS was classified into four scales obtained from damage level of components and structures of NPPs subjected to an earthquake, while EGML was divided into four levels by safe shutdown earthquake ground motion (So), elastic design earthquake ground motion (Sd) and design earthquake ground motion (Ss). Combination of JEDIS and EGML formulated 4 x 4 matrix and determined detailed conditions of restart of NPPs. As a response to an earthquake, operator walk inspections and evaluation of earthquake ground motion were conducted to know the level of JEDIS. JEDIS level requested respective allowable conditions of restart of NPP, which were scale level dependent and consisted of weighted combination of damage inspection (operator walk inspections, focused inspections/tests and expanded inspections), integrity evaluation and repair/replacement. If JEDIS were assigned greater than 3 with expanded inspections, inspection of piping with local wall thinning, its integrity evaluation and repair/replacement if necessary were requested. Inspection and evaluation of piping with local wall thinning was performed based on JSME or ASME codes. Detailed work flow charts were presented. Carbon steel piping and elbow was chosen for evaluation. (T. Tanaka)

Enhanced spatial resolution in fluorescence molecular tomography using restarted L1-regularized nonlinear conjugate gradient algorithm.

Science.gov (United States)

Shi, Junwei; Liu, Fei; Zhang, Guanglei; Luo, Jianwen; Bai, Jing

2014-04-01

Owing to the high degree of scattering of light through tissues, the ill-posedness of fluorescence molecular tomography (FMT) inverse problem causes relatively low spatial resolution in the reconstruction results. Unlike L2 regularization, L1 regularization can preserve the details and reduce the noise effectively. Reconstruction is obtained through a restarted L1 regularization-based nonlinear conjugate gradient (re-L1-NCG) algorithm, which has been proven to be able to increase the computational speed with low memory consumption. The algorithm consists of inner and outer iterations. In the inner iteration, L1-NCG is used to obtain the L1-regularized results. In the outer iteration, the restarted strategy is used to increase the convergence speed of L1-NCG. To demonstrate the performance of re-L1-NCG in terms of spatial resolution, simulation and physical phantom studies with fluorescent targets located with different edge-to-edge distances were carried out. The reconstruction results show that the re-L1-NCG algorithm has the ability to resolve targets with an edge-to-edge distance of 0.1 cm at a depth of 1.5 cm, which is a significant improvement for FMT.

The development of fast reactors in France from March 1980 to March 1981; Le developpement des reacteurs a neutrons rapides en France de mars 1980 a mars 1981

Energy Technology Data Exchange (ETDEWEB)

Vautrey, L. [Commissariat a l' Energie Atomique, CEN de Saclay, Gif-sur-Yvette (France)

1981-05-15

This paper describes general features concerning development in the field of fast reactors in France from March 1980 to March 1981. It concentrates mainly on: Rapsodie, Phenix NPP, prototype reactor Super Phenix 1, future fast reactor NPPs and current research and development programs in the field. The present situation is as follows. Rapsodie has restarted operation but at reduced power in July 1980 because of the problems in the primary circuit which have not yet been solved. Phenic operates in a very satisfactory manner. Construction of Super Phenix is continuing normally. Research activities are performed sometimes for the needs of Super Phenix and sometimes for the needs of future fast rector projects like Super Phenix 2. International cooperation is being continued.

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

International Nuclear Information System (INIS)

2014-12-01

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

A review of the Indian fast reactor programme

International Nuclear Information System (INIS)

Bhoje, S.B.

1990-01-01

Development of Fast Breeder activities is being done mainly at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam and the total Scientific and Technical staff working at the Centre for development of FBRs is about 1200. The development work relating to the fuel fabrication and design and development for some of the fuel handling equipment is being done at the Bhabha Atomic Research Centre, Trombay, Bombay. Complete recovery from the fuel handling incident of FBTR was achieved during the beginning of 1989. Damaged guide tube and bent subassemblies were replaced, the incident was analysed in detail and appropriate remedial measures, viz., modifications in the fuel handling machine control logic and plug rotation logic were implemented to prevent its recurrence. Safety clearances for the restart of the reactor were obtained from the Atomic Energy Regulatory Board in May 1989. As steam generators were not valved in the secondary sodium system, the reactor power during this phase of operation was limited to 500 KWt. The main objectives during this phase were to complete the balance low power physics experiments and to operate the reactor for a sufficiently long time to assess the performance of various systems, in particular the neutronic instrumentation, control rod drive and safety logic system which were not in active service for the two years. From May to July, 1989, the reactor was successfully operated up to a power level of 500 KWt with 50% operating time. Design of PFBR is progressing intensively. (author). 1 tab

Estimation of the sub-criticality of the sodium-cooled fast reactor Monju using the modified neutron source multiplication method

International Nuclear Information System (INIS)

Truchet, G.; Van Rooijen, W. F. G.; Shimazu, Y.; Yamaguchi, K.

2012-01-01

The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and (α, n) reactions which have increased during the shutdown period of 15 years will also be discussed. The relative position of detectors and sources deeply affect the correction factors in some regions. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). (authors)

A New In-core Production Method of Co-60 in CANDU Reactors

Energy Technology Data Exchange (ETDEWEB)

Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

2016-05-15

This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

On the behaviour of dissolved fission gases prior to transient testing of fuel pins

International Nuclear Information System (INIS)

Wood, M.H.; Matthews, J.R.

1978-10-01

The TREAT and CABRI series of reactor safety experiments on irradiated fuel require the transfer of fuel pins from the reactor in which the fuel has achieved some burn-up to the test facility. Subsequently, the fuel is restored to power in the test facility for some time before transient heating is initiated. Such pre-test manoeuvres, where the fuel is subjected to changes in the fission rate and temperature, may have important consequences for the fission gas behaviour during the transient experiment. The results of rate theory calculations are used to assess these effects. (author)

SRS reactor stack plume marking tests

International Nuclear Information System (INIS)

Petry, S.F.

1992-03-01

Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart

Fuel element replacement and cooling water activity at the musashi reactor

International Nuclear Information System (INIS)

Nozaki, Tetsuya; Honda, Teruyuki; Horiuchi, Norikazu; Aizawa, Otohiko; Sato, Tadashi

1989-01-01

The Musashi Institute of Technology Research Reactor (TRIGA 11, 100 kW) has been operated without serious problems since 1963. However, because there is no more spare fuel element, it was necessary to decide how to solve the problem. In the end, it was decided to obtain many stainless steel-clad fuel elements and operate with those fuel elements only, under the auspices of the Ministry of Education, Science and Culture. The bulk shielding experimental pool was remodeled as the storage for spent fuel elements, where the neutrons from the thermalizing column were shielded with cadmium and boron polyethylene plates. The equipment for transferring spent fuel elements was built and temporarily set up between the core tank and the new storage. These works were started in 1983, and finished in 1985. After the reactor was restarted, the count rate of the conventional cooling water monitor which was set in the cooling system using a GM counter drastically decreased. The spent fuel storage, the equipment and the works for fuel transfer, and the radioactivity of cooling water are reported. (K.I.)

Reconstruction for limited-projection fluorescence molecular tomography based on projected restarted conjugate gradient normal residual.

Science.gov (United States)

Cao, Xu; Zhang, Bin; Liu, Fei; Wang, Xin; Bai, Jing

2011-12-01

Limited-projection fluorescence molecular tomography (FMT) can greatly reduce the acquisition time, which is suitable for resolving fast biology processes in vivo but suffers from severe ill-posedness because of the reconstruction using only limited projections. To overcome the severe ill-posedness, we report a reconstruction method based on the projected restarted conjugate gradient normal residual. The reconstruction results of two phantom experiments demonstrate that the proposed method is feasible for limited-projection FMT. © 2011 Optical Society of America

Materials specific work at Forschungszentrum Karlsruhe and in cooperation with the industrial partners ALKEM and Interatom for the development of nuclear oxide fuels for fission reactors

International Nuclear Information System (INIS)

Kleykamp, H.; Muehling, G.

2005-09-01

The fabrication of uranium-plutonium oxide fuel started in Forschungszentrum Karlsruhe and at ALKEM company to begin for the criticality experiments in the SNEAK reactor and subsequently for stationary fuel pin irradiations in the FR2, BR2, DFR, Rapsodie, Phenix and KNK II reactors. The production methods comprised first the mechanical blending of UO2 and PuO2 followed by direct pressing and sintering of the pellets, later the advanced methods such as optimized comilling and ammonium uranyl plutonyl coprecititation. The fabrication of pellets was described in the main, further the alternative fuel pin manufacturing processes by vibrational compaction and hot-impact densification were discussed. The first capsule and pin irradiations in the FR2 and BR2 reactors contributed to the assessment of the maximum operation parameters within the fuel pin development such as linear heat rating, cladding temperature and burnup. Subsequently, small-bundle and largebundle irradiations were made in fast reactors in cooperation with Interatom company in order to verify the specifications for the commercial fast reactor SNR 300. Milestones were the maximum burnup of 175 GWd/t metal, corresponding 18.6 % of the heavy atoms, obtained in one of the KNK II fuel pin assemblies, and the displacement rates in the cladding materials of 140 dpa NRT attained in the Phenix reactor. Higher implications gained later the stationary irradiations of defected mixed-oxide pins, the mild fuel pin transient operations, the local blockage experiments and the severe hypothetic accidents in the respective Siloe, HFR, BR2 and CABRI reactors. These experiments were made solely in international partnership. Further activities were the chemical analyses of solid residues and coprecipitations of irradiated mixed-oxide fuels in the head-end of the reprocessing. All these actions were coordinated in the then fast breeder project. Furthermore, irradiated fuels and fuel pins of other reactor types were

Evaluation of a hybrid anaerobic biofilm reactor treating winery effluents and using grape stalks as biofilm carrier.

Science.gov (United States)

Wahab, Mohamed Ali; Habouzit, Frédéric; Bernet, Nicolas; Jedidi, Naceur; Escudié, Renaud

2016-01-01

Wine production processes generate large amount of both winery wastewater and solid wastes. Furthermore, working periods, volumes and pollution loads greatly vary over the year. Therefore, it is recommended to develop a low-cost treatment technology for the treatment of winery effluents taking into account the variation of the organic loading rate (OLR). Accordingly, we have investigated the sequential operation of an anaerobic biofilm reactor treating winery effluents and using grape stalks (GSs) as biofilm carrier with an OLR ranging from 0.65 to 27 gCOD/L/d. The result showed that, during the start-up with wastewater influent, the chemical oxygen demand (COD) removal rate ranged from 83% to 93% and was about 91% at the end of the start-up period that lasted for 40 days. After 3 months of inactivity period of the reactor (no influent feeding), we have succeeded in restarting-up the reactor in only 15 days with a COD removal of 82% and a low concentration of volatile fatty acids (1 g/L), which confirms the robustness of the reactor. As a consequence, GSs can be used as an efficient carrier support, allowing a fast reactor start-up, while the biofilm conserves its activity during a non-feeding period. The proposed hybrid reactor thus permits to treat both winery effluents and GSs.

Systems aspects of a space nuclear reactor power system

International Nuclear Information System (INIS)

Jaffe, L.; Fujita, T.; Beatty, R.

1988-01-01

Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: Power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, attitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly

Fluid dynamic analysis of a continuous stirred tank reactor for technical optimization of wastewater digestion.

Science.gov (United States)

Hurtado, F J; Kaiser, A S; Zamora, B

2015-03-15

Continuous stirred tank reactors (CSTR) are widely used in wastewater treatment plants to reduce the organic matter and microorganism present in sludge by anaerobic digestion. The present study carries out a numerical analysis of the fluid dynamic behaviour of a CSTR in order to optimize the process energetically. The characterization of the sludge flow inside the digester tank, the residence time distribution and the active volume of the reactor under different criteria are determined. The effects of design and power of the mixing system on the active volume of the CSTR are analyzed. The numerical model is solved under non-steady conditions by examining the evolution of the flow during the stop and restart of the mixing system. An intermittent regime of the mixing system, which kept the active volume between 94% and 99%, is achieved. The results obtained can lead to the eventual energy optimization of the mixing system of the CSTR. Copyright © 2014 Elsevier Ltd. All rights reserved.

Getting to First Flight: Equipping Space Engineers to Break the Start-Stop-Restart Cycle

Science.gov (United States)

Singer, Christopher E.; Dumbacher, Daniel L.

2010-01-01

The National Aeronautics and Space Administration s (NASA s) history is built on a foundation of can-do strength, while pointing to the Saturn/Apollo Moon missions in the 1960s and 1970s as its apex a sentiment that often overshadows the potential that lies ahead. The chronicle of America s civil space agenda is scattered with programs that got off to good starts with adequate resources and vocal political support but that never made it past a certain milestone review, General Accountability Office report, or Congressional budget appropriation. Over the decades since the fielding of the Space Shuttle in the early 1980s, a start-stop-restart cycle has intervened due to many forces. Despite this impediment, the workforce has delivered engineering feats such as the International Space Station and numerous Shuttle and science missions, which reflect a trend in the early days of the Exploration Age that called for massive infrastructure and matching capital allocations. In the new millennium, the aerospace industry must respond to transforming economic climates, the public will, national agendas, and international possibilities relative to scientific exploration beyond Earth s orbit. Two pressing issues - workforce transition and mission success - are intertwined. As this paper will address, U.S. aerospace must confront related workforce development and industrial base issues head on to take space exploration to the next level. This paper also will formulate specific strategies to equip space engineers to move beyond the seemingly constant start-stop-restart mentality to plan and execute flight projects that actually fly.

How to teach geocomplexity at two opposite levels? The 'Klippen of Cabrières' case study (Hérault, France).

Science.gov (United States)

Giusti, Christian

2014-05-01

The starting point of this essay is a question from a school teacher who came in contact with me via my photo gallery online. His question, "What is the origin of the Roc de Murviel?" actually covers three or four separate problems, and requires to devise simple and specific comparisons in order to help children of 8 years old to understand a bunch of difficult and abstract concepts. But, in contrast, Carboniferous marine sedimentary deposits of the so-called unit "Klippen of Cabrières" are known of geologists worldwide, due to the presence of a Global boundary Stratotype Section and Point (GSSP) located not far from the "Roc de Murviel", the "La Serre" section, where is defined the base of the Carboniferous System, Mississippian Sub-System and Tournaisien Stage. In this case, to remind how the conceptions about the formation of the "Klippen of Cabrières" evolved may serve as an introduction to the explanation of syntectonic sedimentation and to the history of the concept of thrust nappe for the well-trained public as well as academics or engineers not experts in the specificities of regional geology. Difficulty (at least in France) will come from the scarcity of naturalists, a species in danger of extinction, and from the small number of historians concerned by the natural sciences. At the public school, the first method is to decompose the problem into independent issues (Cartesian reductionism) such as: nature and age of the rock? mode of sedimentation? part of an ancient mountain chain? existence as a landform in the current landscape? Concerning the specific question of geological time, because most children of this age do not yet know how to read the time, a second method is to replace the model of the clock (the entire history of the planet Earth reduced to one year) by the metaphor of either the staircase or the ladder, both adapted from the international stratigraphic scale. A third method is to use the concept of "toolbox" for each sub-disciplines or

Comparative study of the optimal ratio of biogas production from various organic wastes and weeds for digester/restarted digester

Directory of Open Access Journals (Sweden)

Ugochukwu C. Okonkwo

2018-04-01

Full Text Available This study carried out a comparative analysis of the rates of production of biogas from various organic wastes and weeds which enabled the determination of optimal ratio of poultry droppings to domestic wastes. Digester was prepared for the anaerobic fermentation of the domestic wastes and weeds. The gas production did not begin until the 7th day and increased steadily at first, and then increased sharply until it reached its peak on the 18th day before declining. The total gas produced within the 22 days of experimentation was 1771 cm3. The maximum volume of gas amounting to 809 cm3 was produced by the sample containing 50% poultry dropping and 50% weeds. This indicates that this sample possesses the best C/N ratio of all the samples prepared. For restarted digester, gas production began on the 2nd day as against the 7th day with no restarted digester and the gas production peaked earlier. Keywords: Digester, Optimal ratio, Biogas production, Organic wastes, C/N ratio

A review of fast reactor program in Japan (April 2001 - March 2002)

International Nuclear Information System (INIS)

Nagata, T.; Ieda, Y.

2002-01-01

This report describes the research and development activities on fast reactors in Japan thru April 2001 to March 2002. In December 2001, the Cabinet decided the Plan for Reorganization of Government-funded Corporations including the merger of JNC and the Japan Atomic Energy Research Institute (JAERI). A law to set up a new entity is supposed to be submitted to the National Diet by the Japanese Fiscal Year (JFY) 2004. In the Experimental Fast Reactor Joyo, thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. Though a fire broke out in the maintenance building of Joyo in October 2001, the Mk-III construction work was restarted in February 2002. In the Prototype Fast Breeder Reactor Monju, countermeasures against sodium leakage have already been drawn up based on Monju comprehensive safety review. The safety licensing examination for the plant modification of Monju is undergoing. As for the Feasibility Study on Commercialized Fast Reactor Cycle Systems, JFY2001 was the first year of its second phase. A three-year period from JFY2001 to 2003 is the initial term of this phase. During this term, research activities are being focused on the design of the candidate concepts and fundamental tests of key technologies. An interim summary of these activities will be checked and reviewed, and based on the results; the research for JFY 2004 to 2005 will be conducted in order to narrow down the number of alternatives for the fast reactor cycle. (author)

Interpretation of the CABRI-RAFT RB1 and RB2 tests through detailed data evaluation and PAPAS-2S code analysis

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Sato, Ikken

2001-08-01

The CABRI-RAFT RB1 and RB2 tests were aiming at a study on impact of fuel pin failure under an overpower condition leading to fuel melting. Using a special technique, combination of through-cladding failure and fuel melting was realized. In the RB1 test, fuel ejection was prevented under a limited fuel melting condition. On the other hand, significant fuel melting was applied in the RB2 test so as to get the fuel ejection, thereby obtaining information on the fuel ejection behavior. Interpretation for these tests through the detailed experimental data evaluation and the PAPAS-2S code analysis is performed in this study. Through this study, it is indicated that molten fuel ejection can be prevented with the low smear density fuel as far as the fuel melting is not large for a slit-type cladding defect. Fuel ejection becomes possible in the case of significant fuel melting with a very thin solid fuel shell surrounding the molten fuel cavity. However, the rapidness of the fuel ejection with the low smear density fuel is less pronounced compared with that of the high smear density fuel. It is also confirmed that there is considerable DN-precursor release into the coolant flow already before fuel ejection. The result is very useful for evaluation of anomaly detection with DN signal observation. (author)

Vegetation survey of PEN Branch wetlands

Energy Technology Data Exchange (ETDEWEB)

1991-01-01

A survey was conducted of vegetation along Pen Branch Creek at Savannah River Site (SRS) in support of K-Reactor restart. Plants were identified to species by overstory, understory, shrub, and groundcover strata. Abundance was also characterized and richness and diversity calculated. Based on woody species basal area, the Pen Branch delta was the most impacted, followed by the sections between the reactor and the delta. Species richness for shrub and groundcover strata were also lowest in the delta. No endangered plant species were found. Three upland pine areas were also sampled. In support of K Reactor restart, this report summarizes a study of the wetland vegetation along Pen Branch. Reactor effluent enters Indian Grove Branch and then flows into Pen Branch and the Pen Branch Delta.

Calculation of forces on reactor containment fan cooler piping

International Nuclear Information System (INIS)

Miller, J.S.; Ramsden, K.

2004-01-01

The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology and RELAP5 to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL-96-06 methodology. This evaluation was based on a pressurized water reactor's RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being re-energized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL-96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations. It is shown that both EPRI methodology and RELAP5 calculations can be used to generate hydraulic loads

Research reactor status for future nuclear research in Europe

Energy Technology Data Exchange (ETDEWEB)

Raymond, Patrick; Bignan, Gilles; Guidez, Joel [Commissariat a l' Energie Atomique - CEA (France)

2010-07-01

Scandinavia). The nuclear renaissance is effective worldwide, with 33 power plants today under construction in the world and a lot of projects in discussion or in preparation in various countries (England, Italy, South Africa, USA...). In Europe, some countries, who phase-out the development nuclear energy, are also coming back in nuclear perspectives as Sweden, Italy, England, Poland,.. All these facts begin to give more work to the MTR (material testing reactors) for testing new materials and new fuels to improve their capacities and their performances. For the ZPR (Zero Power Reactors) test with new fuels allowing additives to suppress Bore utilisation, or allowing to reduce uranium consumption, will be necessary in the near future. For the safety dedicated reactors, test for compliance to last safety requirements are necessary. In this field the refurbishment of the CABRI reactor for Reactivity Insertion Accident studies, is now almost finished for test that should begin in 2010. For the radio isotope production the world demand is increasing year after year, especially for {sup 99}Mo, used in about 70 millions of medicine procedures each year in the world. Today 95% of this world production is assumed by five reactors: HFR (Netherlands), OSIRIS (France), SAFARI (South Africa), BRII (Belgium), and NRU (Canada). The youngest is OSIRIS (41 years) and should be close in 2015. Due to ageing problems NRU and HFR were shut down in 2009 for necessary repair. These points have conduced to some radio isotopes crisis in 2009. This paper explains some projects in line for the future to avoid this type of problems (FRMII initiative, RJH utilisation and PALLAS project). For training activities, needs are huge with nuclear renaissance, especially for the new countries coming back in nuclear field. It will also give a lot of opportunities to low power reactors and to the universities reactors. This paper also provides information on the status of the new projects such as the JHR ongoing

Research reactor status for future nuclear research in Europe

International Nuclear Information System (INIS)

Raymond, Patrick; Bignan, Gilles; Guidez, Joel

2010-01-01

renaissance is effective worldwide, with 33 power plants today under construction in the world and a lot of projects in discussion or in preparation in various countries (England, Italy, South Africa, USA...). In Europe, some countries, who phase-out the development nuclear energy, are also coming back in nuclear perspectives as Sweden, Italy, England, Poland,.. All these facts begin to give more work to the MTR (material testing reactors) for testing new materials and new fuels to improve their capacities and their performances. For the ZPR (Zero Power Reactors) test with new fuels allowing additives to suppress Bore utilisation, or allowing to reduce uranium consumption, will be necessary in the near future. For the safety dedicated reactors, test for compliance to last safety requirements are necessary. In this field the refurbishment of the CABRI reactor for Reactivity Insertion Accident studies, is now almost finished for test that should begin in 2010. For the radio isotope production the world demand is increasing year after year, especially for 99 Mo, used in about 70 millions of medicine procedures each year in the world. Today 95% of this world production is assumed by five reactors: HFR (Netherlands), OSIRIS (France), SAFARI (South Africa), BRII (Belgium), and NRU (Canada). The youngest is OSIRIS (41 years) and should be close in 2015. Due to ageing problems NRU and HFR were shut down in 2009 for necessary repair. These points have conduced to some radio isotopes crisis in 2009. This paper explains some projects in line for the future to avoid this type of problems (FRMII initiative, RJH utilisation and PALLAS project). For training activities, needs are huge with nuclear renaissance, especially for the new countries coming back in nuclear field. It will also give a lot of opportunities to low power reactors and to the universities reactors. This paper also provides information on the status of the new projects such as the JHR ongoing construction on the Cadarache

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

International Nuclear Information System (INIS)

1976-09-01

This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format

Moderator behaviour and reactor internals integrity at Atucha I NPP

International Nuclear Information System (INIS)

Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

1996-01-01

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

Moderator behaviour and reactor internals integrity at Atucha I NPP

Energy Technology Data Exchange (ETDEWEB)

Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

1997-12-31

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

Implementation plan for the Waste Experimental Reduction Facility Restart Operational Readiness Review

International Nuclear Information System (INIS)

1993-03-01

The primary technical objective for the WERF Restart Project is to assess, upgrade where necessary, and implement management, documentation, safety, and operation control systems that enable the resumption and continued operation of waste treatment and storage operations in a manner that is compliant with all environment, safety, and quality requirements of the US Department of Energy and Federal and State regulatory agencies. Specific processes that will be resumed at WERF include compaction of low-level compatible waste; size reduction of LLW, metallic and wood waste; incineration of combustible LLW and MLLW; and solidification of low-level and mixed low-level incinerator bottom ash, baghouse fly ash, and compatible sludges and debris. WERF will also provide for the operation of the WWSB which includes storage of MLLW in accordance with Resource Conservation and Recovery Act requirements

Analysis of core damage frequency due to external events at the DOE [Department of Energy] N-Reactor

International Nuclear Information System (INIS)

Lambright, J.A.; Bohn, M.P.; Daniel, S.L.; Baxter, J.T.; Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P.; Brosseau, D.A.

1990-11-01

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs

Analysis of core damage frequency due to external events at the DOE (Department of Energy) N-Reactor

Energy Technology Data Exchange (ETDEWEB)

Lambright, J.A.; Bohn, M.P.; Daniel, S.L. (Sandia National Labs., Albuquerque, NM (USA)); Baxter, J.T. (Westinghouse Hanford Co., Richland, WA (USA)); Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P. (EQE, Inc., San Francisco, CA (USA)); Brosseau, D.A. (ERCE, Inc., Albuquerque, NM (USA))

1990-11-01

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

Directory of Open Access Journals (Sweden)

Yan Wang

2017-01-01

Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

Freeze and restart of the DWPF Scale Glass Melter

International Nuclear Information System (INIS)

Choi, A.S.

1989-01-01

After over two years of successful demonstration of many design and operating concepts of the DWPF Melter system, the last Scale Glass Melter campaign was initiated on 6/9/88 and consisted of two parts; (1) simulation of noble metal buildup and (2) freeze and subsequent restart of the melter under various scenarios. The objectives were to simulate a prolonged power loss to major heating elements and to examine the characteristics of transient melter operations during a startup with a limited supply of lid heat. Experimental results indicate that in case of a total power loss to the lower electrodes such as due to noble metal deposition, spinel crystals will begin to form in the SRL 165 composite waste glass pool in 24 hours. The total lid heater power required to initiate joule heating was the same as that during slurry-feeding. Results of a radiative heat transfer analysis in the plenum indicate that under the identical operating conditions, the startup capabilities of the SGM and the DWPF Melter are quite similar, despite a greater lid heater to melt surface area ratio in the DWPF Melter

Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

International Nuclear Information System (INIS)

Ridha Roslan; Nik Mohd Faiz Khairuddin

2013-01-01

Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

Europe faces up to NSP restart and two new crackers

International Nuclear Information System (INIS)

Roberts, M.

1993-01-01

European cracker operators are hurting-most have not been covering cash costs for the past six months-and they are determined to ease the pain by boosting prices. But since olefins demand remains weak, price gains will have to come via lower production. That appears to be difficult, given the startup of two new world-scale crackers-BP Chemicals (London) 350,000-m.t./year expansion at Grangemouth, UK and EniChem's (Milan) 360,000-m.t./year plant at Brindisi, Italy - and the restart of North Sea Petrochemical's (NSP; Antwerp) 250,000-m.t./year propane dehydrogenation unit. Although the two new crackers have the potential to boost Europe's net olefins output by 4% in 1993, to 18.5 million m.t./year, according to Trichem Consultants (London), the increase will be smaller because EniChem and BP will reduce capacity at other plants as the new units come onstream. EniChem says that the startup of Brindisi will not have an effect on the market. We will not allow the startup to further depress prices

Latest news from the YETS: all restarting except the LHC

CERN Multimedia

Antonella Del Rosso

2016-01-01

With the closure of the SPS at 3 p.m. on Friday, 19 February, maintenance work is now ongoing only at the LHC. All activities are on track for a smooth restart in a few weeks’ time.   At the LHC, all general maintenance activities are proceeding well and according to schedule. In particular, the electrical tests on the general emergency stops have been completed, while the cooling and ventilation maintenance, including leak repairs at various points, will be completed by the end of this week for the whole machine. By the end of next week, the teams will also have completed the bakeout and commissioning of all the collimators, while the installation of coaxial cable for clock distribution for CMS TOTEM has been postponed to the Extended Year-End Technical Stop (EYETS), scheduled to start in December. Following a recent decision, additional electrical tests of the circuits (ELQA and energy extraction insulation tests) for the whole machine have been added to the schedule. The SPS is currentl...

Spent fuel from RA reactor inspection of state and options for management

International Nuclear Information System (INIS)

Aden, V.G.; Bulkin, S. Yu.; Sokolov, A. V.; Matausek, M.V.; Vukadin, Z.

2001-01-01

About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. The realization of this technology was started in 1999 but due to political and financial difficulties was not completed. In September the year 2000 the work was restarted. Two different ways of RA reactor spent fuel elements preparation for transportation or long-term storage are considered: 'all fuel elements canning without leak-tightness testing' and 'all fuel elements leak-tightness testing'. It is believed that the first option offers several distinct advantages, which can be summarized as: greater reliability in the course of transportation or dry storage. Higher safety for workers. Lower expenditures for non-standard equipment manufacturing. Shorter duration of work. (author)

Restart 2.0 of substrategic nuclear weapon disarmament? Negotiation approaches and models; Neustart 2.0 zur Abruestung substrategischer Nuklearwaffen? Verhandlungsansaetze und -modelle

Energy Technology Data Exchange (ETDEWEB)

Paul, Michael

2011-05-15

The author discusses the following topics with respect a possible restart of nuclear disarmament negotiations: nuclear disarmament versus nuclear deterrence; substrategic nuclear weapons; initial positions for the negotiations (American position, Russian position, German position); strategic and substrategic nuclear weapon disarmament (including the questions of transparency and verification); imponderables.

SP-100 initial startup and restart control strategy

International Nuclear Information System (INIS)

Halfen, F.J.; Wong, K.K.; Switick, D.M.; Shukla, J.N.

1992-01-01

This paper reports that recent Generic Flight System (GFS) updates have necessitated revisions in the initial startup and restart control strategies. The design changes that have had the most impact on the control strategies are the addition of the Auxiliary Cooling and Thaw (ACT) system for preheating the lithium filled components, changes in the reactivity worths of the reflectors and safety-rods such that initial cold criticality is achieved with only a small amount of reflector movement following the withdrawal of the safety-rods, and the removal of the scram function from the reflectors. Revised control and operating strategies have been developed and tested using the SP-100 dynamic simulation model, ARIES-GFS. The change in the total reactivity worths of the reflectors and safety-rods has eliminated the need for the use of fast and slow reflector drive speeds during the initial on-orbit approach to criticality. The relatively fast removal of the safety-rods results in a near-critical condition so that the use of slow moving (single speed) reflector drives does not add significant time to achieve full power for the initial startup. The use of the ACT system (with its NaK trace-lines for preheating and auxiliary cooling) affects the main Thermoelectric Electro-Magnetic (TEM) pump startup and the time after a shutdown before freezing occurs in the main heat transfer systems

HFBR restart activity A2.6: Review of FSAR and 60 MW addendum to assure consistency of operation at 40 MW

International Nuclear Information System (INIS)

Rao, D.V.; Ross, S.B.; Darby, J.L.; Clark, R.A.

1990-01-01

The purpose of this task (HFBR Restart Activity A2.6) is to perform a review of the design basis accident (DBA) analyses sections of the 1964 HFBR-Final Safety Analysis Report; Volumes I and II, and the 1982 Addendum to the HFBR-FSAR for 60 MW operation to assure that operation at 40 MW will be consistent with these analyses. Additional documents utilized in the review included the Level 1 PRA for HFBR, HFBR-PDMs and HFBR-OPMs. The review indicates that the 1964 FSAR-DBA analysis in incomplete in the sense that it did not analyze some of the important initiators for 1-loop operation that include: Accidental throttling of primary flow control valves; seizure of primary pump; loss of secondary pump; accidental throttling of secondary flow control valves; rupture of secondary piping. The first three initiators were later studied in the 1982 addendum. The other two initiators have not been examined to-date for 1-loop operation. It is recommended that the impact of these initiators be assessed prior to the restart, if 1-loop operation is chosen for the restart. The review demonstrated that at 40 MW operation there are only a few accident initiators that will culminate in core damage (fuel melting and /or cladding failure) regardless of the availability of mitigating systems. For 1-loop Operation these accidents include: Fuel channel blockage, primary pump seizure, and large-large LOCA (a LOCA with effective break diameter > 2.81 in. is referred to as a large-large LOCA in this document as well as in PRA). Although all these accidents listed above could lead to core damage for 1-loop operation as well, the probability is expected be very low

FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

Energy Technology Data Exchange (ETDEWEB)

GANTT, D.A.

2000-10-31

This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

Asynchronous Checkpoint Migration with MRNet in the Scalable Checkpoint / Restart Library

Energy Technology Data Exchange (ETDEWEB)

Mohror, K; Moody, A; de Supinski, B R

2012-03-20

Applications running on today's supercomputers tolerate failures by periodically saving their state in checkpoint files on stable storage, such as a parallel file system. Although this approach is simple, the overhead of writing the checkpoints can be prohibitive, especially for large-scale jobs. In this paper, we present initial results of an enhancement to our Scalable Checkpoint/Restart Library (SCR). We employ MRNet, a tree-based overlay network library, to transfer checkpoints from the compute nodes to the parallel file system asynchronously. This enhancement increases application efficiency by removing the need for an application to block while checkpoints are transferred to the parallel file system. We show that the integration of SCR with MRNet can reduce the time spent in I/O operations by as much as 15x. However, our experiments exposed new scalability issues with our initial implementation. We discuss the sources of the scalability problems and our plans to address them.

Neutron calculation scheme for coupled reactors

International Nuclear Information System (INIS)

Porta, Jacques.

1980-11-01

The CABRI and PHEBUS cores are of the low enrichment rod type in which the fuel is made up of uranium oxide pellets encased in tubular cladding but the SCARABEE core has high enrichment plates, the fuel, an aluminium-uranium alloy (UAl) is metal, rolled into plate form. These three cores in well described rectangular geometry, receive in their centres the very heterogeneous cylindrical test loops (numerous containments of different kinds, large void spaces acting as lagging). After a detailed study of these three reactors, it is found that the search for a calculation scheme (common to the three projects) leads to the elimination of the scattering approximation in our calculations. It is therefore necessary to review the various existing models from a theoretical angle and then to select a particular method, according to the available data processing tools, a choice that will be dictated by the optimization of the parameters: cost in calculation time, difficulties (or ease) of use and accuracy achieved. A problem of experiment interpretation by calculation is dealt with in Chapter 3. The determination of the coupling by calculation is closely linked to the geometrical and energy modelization chosen. But from the experimental angle the determination of the coupling also gives rise to problems with respect to the interpretation of the experimental values obtained by thermal balance determinations, counting of the gamma emission of the fission products of fissile detectors and counting of lanthane 140 in the fuel fission products. The method of calculation is discussed as is the use made of detectors and the counting procedures. In chapter 4, it is not a local modelization that is discussed but an overall one in an original three dimensional calculation [fr

Protection of biofilms against toxic shocks by the adsorption and desorption capacity of carriers in anaerobic fluidized bed reactors

Energy Technology Data Exchange (ETDEWEB)

Petrozzi, S. (Biological Reaction Engineering Group, Chemical Engineering Dept., ETH, Zurich (Switzerland)); Kut, O.M. (Biological Reaction Engineering Group, Chemical Engineering Dept., ETH, Zurich (Switzerland)); Dunn, I.J. (Biological Reaction Engineering Group, Chemical Engineering Dept., ETH, Zurich (Switzerland))

1993-05-01

The aim of this study was to select a support medium for an anaerobic biofilm fluidized bed reactor (AFBR) for waste water treatment. Six materials, shale, pumice, porous glass, quartz sand, activated carbon and anthracite were used as carriers for the biofilm. The reactors were operated in parallel for several months with vapour condensate from a sulfite cellulose process as feed. The criteria used for the evaluation were: (a) Reproducibility of the reactor performance, (b) performance of the different carriers under various loading rates, (c) stability against toxic shock loadings using 2,4,6-trichlorophenol (TCP) as toxicant, (d) recovery capacity after intoxication and starvation, (e) adsorption/desorption behavior of the carriers. A comparison between four runs showed good reproducibility of the steady state removal rates. The performance of the reactors and the stability of the degradation rates were tested for a range of loading conditions. Unbuffered, buffered and pH controlled conditions were compared. The pumice carrier was best with respect to the degradation rate achieved per carrier mass. The response of the reactors to massive TCP step loadings was tested. Loadings less than 1.5 kg TCP/m[sup 3]d resulted in initially normal gas production rates for all the systems, except the activated carbon, whose gas production was partially inhibited from the start. After increasing the load to 1.5 kg TCP/m[sup 3]d the gas production rates of all the other reactors fell abruptly to zero. Restarting after 2 months, all reactors showed methanogenic activity without requiring new inoculum. (orig.)

Efficient L1 regularization-based reconstruction for fluorescent molecular tomography using restarted nonlinear conjugate gradient.

Science.gov (United States)

Shi, Junwei; Zhang, Bin; Liu, Fei; Luo, Jianwen; Bai, Jing

2013-09-15

For the ill-posed fluorescent molecular tomography (FMT) inverse problem, the L1 regularization can protect the high-frequency information like edges while effectively reduce the image noise. However, the state-of-the-art L1 regularization-based algorithms for FMT reconstruction are expensive in memory, especially for large-scale problems. An efficient L1 regularization-based reconstruction algorithm based on nonlinear conjugate gradient with restarted strategy is proposed to increase the computational speed with low memory consumption. The reconstruction results from phantom experiments demonstrate that the proposed algorithm can obtain high spatial resolution and high signal-to-noise ratio, as well as high localization accuracy for fluorescence targets.

Results of environmental radiation monitoring and meteorology measurements (material prepared for obtaining the licence for RA reactor experimental operation); Rezultati merenja zracenja u okolini i rezultati meteoroloskih merenja (materijal pripremljen radi dobijanja dozvole za pustanje Reaktora RA u probni rad)

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-10-15

According to the demands for obtaining the licence for restarting the Ra reactor and the experimental operation this document includes the radiation monitoring measured data in the working space and environment of the RA reactor, i.e. Boris Kidric Institute. The meteorology measured data are included as well. All the measurements are performed according to the radiation protection program applied actually from the first reactor start-up at the end of 1959. [Serbo-Croat] Saglasno zahtevu za dobijanje dozvole za ponovno pustanje u probni rad reaktora RA, ovaj dokument sadrzi rezultate merenja zracenja u okolini (radnoj i zivotnoj) reaktora RA odnosno instituta 'Boris Kidric' kao i podatke o meteoroloskim merenjima. Sve merenja rade se prema programu mera zastite od zracenja koje se sprovode prakticno od prvog pustanja reaktora u rad krajem 1959. godine.

Interpretation of the CABRI-RAFT LTX test up to pin failure based on detailed data evaluation and PARAS-2S code analysis

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Sato, Ikken

2001-09-01

The CABRI-RAFT LTX test aims at a study on the fuel-pin-failure mechanism, in-pin fuel motion and post-failure fuel relocation with an annular fuel pin which was pre-irradiated up to peak burn-up of 6.4 at%. The transient test conditions similar to those of the LT4 test were selected in the LTX test using the same type of fuel pin, allowing an effective direct comparison between the two tests. In contrast to the LT4 test which showed a large PCMI-mitigation potential of the annular fuel-pin design, early pin failure occurred in the LTX test when fuel does not seem to have molten. In order to clarify the fuel pin failure mechanism, interpretation of the LTX test up to pin failure is performed in this study, through an experimental data evaluation and a PAPAS-2S-code analysis. The PAPAS-2S code simulates reasonably the fuel thermal conditions such as transient fuel-pin heat-up and fuel melting. The present detailed data evaluation shows that the earlier cladding failure compared with the LT4 test is mainly attributed to the local cladding heat-up. Under the high-temperature condition, plenum gas pressure has a certain potential to explain the observed failure. Fuel swelling-induced PCMI does not seem significant in the LTX test and it may have contributed to the early pin failure only to a limited extent, if any. (author)

Safety Evaluation Report related to the restart of Rancho Seco Nuclear Generating Station, Unit 1, following the event of December 26, 1985 (Docket No. 50-312)

International Nuclear Information System (INIS)

1987-10-01

On December 26, 1985, the Rancho Seco Nuclear Generating Station, owned and operated by the Sacramento Municipal Utility District (SMUD), experienced a loss of dc power within the integrated control system (ICS) while the plant was at 76% power. The ensuing reactor trip was followed by a rapid overcooling transient and automatic initiation of the safety features actuation system (SFAS). The overcooling transient continued until ICS dc power was restored 26 minutes after its loss. Two letters from the NRC Region V Administrator (dated December 26, 1985) confirmed that the Rancho Seco plant would not be returned to power operation until SMUD (the licensee) had provided the NRC with an assessment of the root cause of the transient and a justification as to why the Rancho Seco facility is ready to resume power operation. In response, the licensee submitted the ''Rancho Seco Action Plan for Performance Improvement'' on July 3, 1986; revisions to that action plan were submitted on December 15, 1986 and February 28, 1987. The NRC staff has reviewed the action plan and numerous other supporting documents submitted by the licensee. The staff's evaluation of the information supporting restart of Rancho Seco is presented in this safety evaluation report

Creys-Melville: A case study on the acceptability of fast breeder reactors

Energy Technology Data Exchange (ETDEWEB)

Lacroix, Andre; D' Heilly, Brigitte [NERSA-EdF (France)

1993-07-01

The Operator of CREYS-MALVILLE, being convinced that in a democratic country like France, Nuclear Energy in general and fast reactors in particular may only be developed in agreement with public opinion, will continue to develop its external Communication - without passion but with conviction. This paper has recalled the interests at stake associated with the resumption of power at CREYS-MALVILLE, the media and political contexts which have followed the construction and operation of this Plant, the administrative resources, and the communication equipment which have been developed over these years. All these means would be wasted and all communication efforts useless if they did not rely on the perseverance and competence of the 650 technicians and engineers guarantors of the Safety at CREYS-MALVILLE, for whom the coming Public Enquiry will be yet another occasion to explain, to France and its neighbours, the necessity to restart SUPERPHENIX.

Creys-Melville: A case study on the acceptability of fast breeder reactors

International Nuclear Information System (INIS)

Lacroix, Andre; D'Heilly, Brigitte

1993-01-01

The Operator of CREYS-MALVILLE, being convinced that in a democratic country like France, Nuclear Energy in general and fast reactors in particular may only be developed in agreement with public opinion, will continue to develop its external Communication - without passion but with conviction. This paper has recalled the interests at stake associated with the resumption of power at CREYS-MALVILLE, the media and political contexts which have followed the construction and operation of this Plant, the administrative resources, and the communication equipment which have been developed over these years. All these means would be wasted and all communication efforts useless if they did not rely on the perseverance and competence of the 650 technicians and engineers guarantors of the Safety at CREYS-MALVILLE, for whom the coming Public Enquiry will be yet another occasion to explain, to France and its neighbours, the necessity to restart SUPERPHENIX

RA Research reactor, Annual report 1970 - Operation and maintenance

International Nuclear Information System (INIS)

Milosevic, D. et al.

1970-12-01

During 1970, the RA Reactor was operated at nominal power of 6.5 MW for 160 days, and 40 days at lower power levels. Total production mounted to 25968 MWh which is 3.87% higher than planned. The action plan was changed compared to the previous years because of sending the heavy water to France for re-concentration. Isotopic concentration of the heavy water was decreased to 99.05% and now after re-concentration it is 99.96%. Discrepancy from the action plan, in September was caused by the delay return of the heavy water for administrative and transportation difficulties. The restart of the reactor in September was postponed because the cladding of one fuel element was damaged immediately after the start-up, and the reactor had to be shutdown. In October and November reactor was in operation 28 and 25 days respectively which enabled to make up for the lost time. Reactor was used for irradiation and experiments according to the demand of 390 users, 340 from the Institute and 50 external users. This report contains detailed data about reactor power and experiments performed in 1969. It is concluded that the reactor operated successfully according to the plan. Shorter interruptions were caused only by difficulties with water supply pipes and sliding of the soil. Reactor was only twice scram shutdown because of the false signals caused by failures of the electronic control instrumentation. the period when reactor was not in operation was used for inspection of the reactor vessel internals. By using special TV cameras and telescopes, it was found that the there are no signs of corrosion on the reactor vessel, e.e. that the internals are in a very good state. Simultaneously, connection for the pipes of future emergency core cooling system were constructed. During 1970, the spent fuel was repacked from fuel channels into special aluminium casks. Four casks containing 660 fuel slugs was deposited int the storage pool No.4. There is now 18 casks with 2951 spent fuel slugs in

Fusion reactors-high temperature electrolysis (HTE)

International Nuclear Information System (INIS)

Fillo, J.A.

1978-01-01

Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 1800 0 C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400 0 C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) [$1000/KW(E) equivalent], the H 2 energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 10 6 scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen

APPLICATION OF RESTART COVARIANCE MATRIX ADAPTATION EVOLUTION STRATEGY (RCMA-ES TO GENERATION EXPANSION PLANNING PROBLEM

Directory of Open Access Journals (Sweden)

K. Karthikeyan

2012-10-01

Full Text Available This paper describes the application of an evolutionary algorithm, Restart Covariance Matrix Adaptation Evolution Strategy (RCMA-ES to the Generation Expansion Planning (GEP problem. RCMA-ES is a class of continuous Evolutionary Algorithm (EA derived from the concept of self-adaptation in evolution strategies, which adapts the covariance matrix of a multivariate normal search distribution. The original GEP problem is modified by incorporating Virtual Mapping Procedure (VMP. The GEP problem of a synthetic test systems for 6-year, 14-year and 24-year planning horizons having five types of candidate units is considered. Two different constraint-handling methods are incorporated and impact of each method has been compared. In addition, comparison and validation has also made with dynamic programming method.

RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1981-12-15

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

Life stage differences in resident coping with restart of the Three Mile Island nuclear generating facility

International Nuclear Information System (INIS)

Prince-Embury, S.; Rooney, J.F.

1990-01-01

A study of residents who remained in the vicinity of Three Mile Island (TMI) immediately following the restart of the nuclear generating plant revealed that older residents employed a more emotion-focused coping style in the face of this event than did younger residents. Coping style was, however, unrelated to the level of psychological symptoms for these older residents, whereas demographic variables were related. Among younger residents, on the other hand, coping style was related to the level of psychological symptoms, whereas demographic variables were not. Among younger residents, emotion-focused coping was associated with more symptoms and problem-focused coping was associated with fewer symptoms, contradicting previous findings among TMI area residents

Development of a MELCOR self-initialization algorithm for boiling water reactors

International Nuclear Information System (INIS)

Chien, C.S.; Wang, S.J.; Cheng, S.K.

1996-01-01

The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users' manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR users an easy and correct method for establishing the initial conditions

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

Energy Technology Data Exchange (ETDEWEB)

Diamond, David

2017-12-29

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allow for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.

Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

significantly lower enthalpies than those required for melting, when the fuel burnup exceeds approximately 40 MWd/kgU. This issue is investigated by reviewing all high-burnup UO 2 fuel rods that have failed in RIA simulation tests in the Japanese Nuclear Safety Research Reactor and the French CABRI pulse reactor to date. Data from thirteen failed rods, with burnups between 44 and 64 MWd/kgU, indicate that clad tube failure does not necessarily lead to fuel pellet dispersal. In fact, the data suggest that a peak fuel radial average enthalpy of at least 500 J/gUO 2 is required to expel a significant part (>10 %) of the fuel inventory into the coolant. However, this empirical enthalpy threshold for fuel dispersal from high-burnup fuel rods cannot be directly applied to light water reactors, since the power pulses and/or the cooling conditions used in the pulse tests differ notably from those expected in LWRs under RIA

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988; Istrazivacki nuklearni reaktor RA, deo 1, pogon i odrzavanje nukleanog reaktora RA u 1988. godini

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1988-12-15

According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues. [Serbo-Croat] Prema planu za 1988. godinu, reaktor RA je trebalo da pusten u rad oktobra meseca, medjutim nije dobio dozvolu za nastavak rada. Kontrola i odrzavanje opreme izvrsavani su redovno i efikasno, u granicama koje su diktirane raspolozivoscu repromaterijala i rezervnih delova. Najvecu poteskocu pricinjavalo je odrzavanje instrumentacije. Period stajanja u 1988. godini iskoriscen je za remont teskovodnih pumpi u primarnom kolu hladjenja. U cilju povecanja pouzdanosti rada reaktora zapoceti su radovi na modernizaciji instrumentacije, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, sklopljen je ugovor o izradi ove opreme koja bi trebalo da bude isporucena do kraja 1990. U cilju povecanja prostora za skladistenje ozracenog

Restart oversight assessment of Hanford 242-A evaporator: Technical report

International Nuclear Information System (INIS)

Lagdon, R.; Lasky, R.

1994-08-01

An assessment team from the Office of Environment, Safety and Health (EH), US Department of Energy (DOE), conducted an independent assessment of the 242-A Evaporator at the Hanford Site during January 17--28, 1994. An EH team member remained on-site following the assessment to track corrective actions and resolve prestart findings. The primary objective of this assessment was independent assurance that the DOE Office of Environmental Management (EM), the DOE Richland Operations Office (DOE-RL), and Westinghouse Hanford Company (WHC) can safely restart the evaporator. Another objective of the EH team was to assess EM's Operational Readiness Evaluation (ORE) to determine if the programs, procedures, and management systems implemented for operation of the 241-A Evaporator ensure the protection of worker safety and health. The following section of this report provides background information on the 242-A Evaporator and Operational Readiness Review (ORR) activities conducted to date. The next chapter is divided into sections that address the results of discrete assessment activities. Each section includes a brief statement of conclusions for the functional area in question, descriptions of the review bases and methods, and a detailed discussion of the results. Concerns identified during the assessment are listed for the section to which they apply, and the specific findings upon which the concern is based can be found immediately thereafter

SIMIFR: A code to simulate material movement in the Integral Fast Reactor

International Nuclear Information System (INIS)

White, A.M.; Orechwa, Yuri.

1991-01-01

The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

Investigating the Problem Solving Competency of Pre Service Teachers in Dynamic Geometry Environment

Science.gov (United States)

Haja, Shajahan

2005-01-01

This study investigated the problem-solving competency of four secondary pre service teachers (PSTs) of University of London as they explored geometry problems in dynamic geometry environment (DGE) in 2004. A constructivist experiment was designed in which dynamic software Cabri-Geometre II (hereafter Cabri) was used as an interactive medium.…

A probabilistic analysis of rapid boron dilution scenarios

International Nuclear Information System (INIS)

Kohut, P.; Diamond, D.J.

1993-01-01

A probabilistic and deterministic analysis of a rapid boron dilution scenario related to reactor restart was performed. The event is initiated by a loss of off-site power during the startup dilution process. The automatic restart of the charging pump in such cases may lead to the accumulation of a diluted slug of water in the lower plenum. The restart of the reactor coolant pumps may send the diluted slug through the core, adding sufficient reactivity to overcome the shutdown margin and cause a power excursion. The concern is that the power excursion is sufficient in certain circumstances to cause fuel damage. The estimated core damage frequency based on the scoping analysis is 1.0--3.0E-05/yr for the plants analyzed. These are relatively significant values when compared to desirable goals. The analysis contained assumptions related to plant specific design characteristics which may lead to non-conservative estimates. The most important conservative assumptions were that mixing of the injected diluted water is insignificant and that fuel damage occurs when the slug passes through the core

Refurbishment and safety management of JMTR in extended showdown

International Nuclear Information System (INIS)

Ide, Hiroshi; Hori, Naohiko; Gorai, Shigeru; Kusunoki, Tsuyoshi

2011-06-01

Japan Materials Testing Reactor (JMTR) is a testing reactor dedicated to the irradiation tests of materials and fuels. The reactor type of the JMTR is light water moderated and cooled tank type. It achieved first criticality in 1968. Operation was started in 1970. The JMTR had been being operated for 38 years from first criticality to the JMTR No.165 cycle finished. Periodic Safety Review (PSR) was carried out with confirming the integrity inspection of the JMTR reactor facilities. And the 10 years maintenance plan was made in 2004. After that, the restart of the JMTR has been strongly requested from various users as the only irradiation testing reactor in Japan. Finally, Japan Atomic Energy Agency (JAEA) decided the refurbishment and restart of the JMTR in December 2006, and the refurbishment works was started from FY 2007. The equipment to remain in use and that which needs replacing before the restart of the JMTR was selected after having been evaluated on its damage and wear due to aging significance in safety functions, past safety-related maintenance date, and the enhancement of facility operation. The renewal work of power supply system, boiler, radioactive waste facility, etc. was already carried out as scheduled. The renewal work of reactor control system, nuclear instrumentation system and so on is being carried out. As for the safety management during reactor operation, the facility periodical own inspection and daily inspection is carried out for the purpose of maintaining soundness and reliability of facilities and equipments. And it is confirmed that the performance of facilities and equipments is maintained. As for the radiation control, irradiation dose limit determined by the law is obeyed. Based on the Concept of radiation protection of the International Commission on Radiation Protection (ICRP), reduction of dose is endeavored. The safety management during reactor shutdown is also carried out as well as it of reactor operation term. However, the

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

Energy Technology Data Exchange (ETDEWEB)

Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-11-30

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

Plenary session. Current status of JMTR

International Nuclear Information System (INIS)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Komori, Yoshihiro; Suzuki, Masahide

2012-01-01

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and the check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Stress test, what is the reality and significance of it?

International Nuclear Information System (INIS)

Sawada, Tetsuo

2012-01-01

Stress test was introduced in July 2011 by 'political judgment' to demonstrate the ability of nuclear power plants to withstand severe earthquake and tsunami. Stress test consisted of two stages and the first stage using computerized simulation required to obtain 'cliff edge' for earthquake, tsunami, their superposition, loss of all alternating current power and loss of final heat sink, and effectiveness of severe accident management after emergency safety measures. Clearing the first stage of the test was a prerequisite for restarting reactors that had been suspended for regular inspections. NISA had received such test results for 14 nuclear reactors as of January 18, 2012. After passing IAEA's evaluation of stress test review process, NISA's endorsement of test results, NSC's confirmation of NISA's screening results and approval of local government, Prime Minister and relevant ministers concerned would decide whether reactors could be restarted as 'political judgment'. Using ranking list and referring to respective experiences of 14 reactors hit by earthquake and tsunami at the Great East Japan earthquake might better perform comprehensive judgment. (T. Tanaka)

Steel Creek water quality: L-Lake/Steel Creek Biological Monitoring Program, November 1985--December 1991

International Nuclear Information System (INIS)

Bowers, J.A.; Kretchmer, D.W.; Chimney, M.J.

1992-04-01

The Savannah River Site (SRS) encompasses 300 sq mi of the Atlantic Coastal Plain in west-central South Carolina. The Savannah River forms the western boundary of the site. Five major tributaries of the Savannah River -- upper Three Runs Creek, Four Mile Creek, Pen Branch, Steel Creek, and Lower Three Runs Creek -- drain the site. All but Upper Three Runs Creek receive, or in the past received, thermal effluents from nuclear production reactors. In 1985, L Lake, a 400-hectare cooling reservoir, was built on the upper reaches of Steel Creek to receive effluent from the restart of L-Reactor, and protect the lower reaches from thermal impacts. The Steel Creek Biological Monitoring Program was designed to meet envirorunental regulatory requirements associated with the restart of L-Reactor and complements the Biological Monitoring Program for L Lake. This extensive program was implemented to address portions of Section 316(a) of the Clean Water Act. The Department of Energy (DOE) must demonstrate that the operation of L-Reactor will not significantly alter the established aquatic ecosystems

Third session: new projects

International Nuclear Information System (INIS)

Iracane, D.; Dupuy, J.P.; Perotto, G.; Ithurralde, G.; Cabrillat, J.C.; Becle, D.; Pivet, S.; Minguet, J.L.

2005-01-01

The RJH (Jules Horowitz reactor) is a 100 MWth research reactor, cooled and moderated by water, that has been designed to operate with high density (8 g U/cm 3 ) and low enrichment fuel (19.75%) uranium molybdenum fuel. Its main neutron performances are expected to be: -) maximum fast neutron flux (E > 0.907 MeV) in the sample: 5.10 14 n/cm 2 /s, -) maximum thermal neutron flux (E 14 n/cm 2 /s, and -) maximal neutron damage: 16 dpa/year. RJH that will be built on the Cea's site of Cadarache, is scheduled to enter into service in 2014, it will take over the activities of the Osiris reactor and its operating life is planned to be over 50 years long. The Cabri research reactor is located on the Cadarache site, it has been operating since 1962 and is now undergoing important upgrading operations whose aim is to make the facility fit to study the behaviour of new nuclear fuels in accidental situations. The RES test reactor is a facility whose aim is to qualify nuclear fuels and core configurations of existing propulsion reactors onboard of aircraft carriers or submarines and to propose innovative concepts for future propulsion reactors. (A.C.)

Analogical study of the servo-control of a reactivity modulator

International Nuclear Information System (INIS)

Le Bot, Michel

1969-03-01

In the context of the study of the transfer functions related to the Cabri reactor, this paper presents: the objective of the servo-control (reactivity modulator, reasons for the analogical study), the principles of the servo-control (description of the servo-controlled system, elaboration of the error signal, principles of the phase meter, critical analysis of different types of phase meters), the analogical formulation (transfer diagram of the process, analogical simulation of the process, analogical realization of the phase meter, simulation of the Low Frequency generator), study of the controlled system and results (system responses to echelon and ramp signals, responses of the controlled system with the phase meter in feedback)

Fast Inbound Top-K Query for Random Walk with Restart.

Science.gov (United States)

Zhang, Chao; Jiang, Shan; Chen, Yucheng; Sun, Yidan; Han, Jiawei

2015-09-01

Random walk with restart (RWR) is widely recognized as one of the most important node proximity measures for graphs, as it captures the holistic graph structure and is robust to noise in the graph. In this paper, we study a novel query based on the RWR measure, called the inbound top-k (Ink) query. Given a query node q and a number k , the Ink query aims at retrieving k nodes in the graph that have the largest weighted RWR scores to q . Ink queries can be highly useful for various applications such as traffic scheduling, disease treatment, and targeted advertising. Nevertheless, none of the existing RWR computation techniques can accurately and efficiently process the Ink query in large graphs. We propose two algorithms, namely Squeeze and Ripple, both of which can accurately answer the Ink query in a fast and incremental manner. To identify the top- k nodes, Squeeze iteratively performs matrix-vector multiplication and estimates the lower and upper bounds for all the nodes in the graph. Ripple employs a more aggressive strategy by only estimating the RWR scores for the nodes falling in the vicinity of q , the nodes outside the vicinity do not need to be evaluated because their RWR scores are propagated from the boundary of the vicinity and thus upper bounded. Ripple incrementally expands the vicinity until the top- k result set can be obtained. Our extensive experiments on real-life graph data sets show that Ink queries can retrieve interesting results, and the proposed algorithms are orders of magnitude faster than state-of-the-art method.

Production of sealed sup 6 sup 0 Co and sup 1 sup 9 sup 2 Ir sources of high specific activity in the nuclear reactor RA

International Nuclear Information System (INIS)

Dobrijevic, R.; Vucina, J.

1998-01-01

Given is a review on the development of the production of 60 Co and 192 Ir performed in the Vinca Institute in the nuclear reactor RA. The experience gained showed that this reactor was suitable for obtaining of these and some other radionuclides. One possibility of its re-start is that the performances of the reactor remain the same (power 6.5 MW, max.neutron flux up to 6x10 13 n.cm -2 s -1 ). By applying new techniques of target preparation, 60 Co for sterilization units of specific activity 1.11 TBq/g could be produced. Maximal activity of sup 1 sup 9 sup 2 Ir would be about 1.48 TBq what is satisfactory for the sources for gamma radiography. The increase of the flux to 10 14 n.cm -2 s -1 would enable the production of 60 Co of specific activities about 3.335 TBq/g. This is satisfactory for the sources for the radiation therapy of activities up to 111 TBq and for gamma radiography of activities about 0.37 TBq. In the case of 192 Ir the sources for the radiation therapy of activities about 0.37 TBq could be obtained. Maximal achievable activities of 192 Ir would be about 3.7 TBq. (author)

Heat exchanger restart evaluation

International Nuclear Information System (INIS)

Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

1992-01-01

On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4kA was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized herein

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Directory of Open Access Journals (Sweden)

Isabelle Guénot-Delahaie

2018-03-01

Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

Heat exchanger restart evaluation

International Nuclear Information System (INIS)

Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

1992-01-01

On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized

Heat exchanger restart evaluation

International Nuclear Information System (INIS)

Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

1992-01-01

On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summary herein

GPU credit reduced, tie to TMI-1 cheating discounted

International Nuclear Information System (INIS)

Utroska, D.

1981-01-01

The recent reduction of credit available to General Public Utilities (GPU) Nuclear may be linked to a cheating incident involving two reactor operators at the Three Mile Island-1 (TMI-1) reactor. The incident caused the Nuclear Regulatory Commission to reopen the managerial portion of the restart hearings and may delay the restart. The delay and the lower credit line will worsen GPU's financial position. Banks claim that misgivings about TMI-1 influence them more than the cheating, although GPU had been gradually improving its financial situation since the TMI-2 accident. The new agreement gives GPU $150 million in immediate credit, but lowers the interim ceiling from $292 million to $200 million. A spokesman from the Office of Management and Budget acknowledges that administration plans to limit the federal role to research and development softened under political pressure

Progress report I.R.S.N. 2003; Rapport d'activite IRSN 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

This report is divided in six chapters. The first one is devoted to the safety of civil installations, transport and radioactive wastes. In this chapter we find the safety of PWR type reactors, the safety of Cabri reactor, ITER tokamak, the storage of radioactive waste and the criticality of packages with actinides. The second chapter concerns the protection of man and environment, we find the radiation monitoring of workers, the medical management of irradiated personnel, the radiation protection of nuclear power plants. The third chapter is relative to the national defense. The fourth chapter tackles the international relations and the part taken by I.R.S.N. in the Europe construction. The fifth chapter gives the contribution to public information, communication and training. The sixth and last chapter explains the quality policy at I.R.S.N. and how the management is made by a quality option. (N.C.)

Progress report I.R.S.N. 2003; Rapport d'activite IRSN 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

This report is divided in six chapters. The first one is devoted to the safety of civil installations, transport and radioactive wastes. In this chapter we find the safety of PWR type reactors, the safety of Cabri reactor, ITER tokamak, the storage of radioactive waste and the criticality of packages with actinides. The second chapter concerns the protection of man and environment, we find the radiation monitoring of workers, the medical management of irradiated personnel, the radiation protection of nuclear power plants. The third chapter is relative to the national defense. The fourth chapter tackles the international relations and the part taken by I.R.S.N. in the Europe construction. The fifth chapter gives the contribution to public information, communication and training. The sixth and last chapter explains the quality policy at I.R.S.N. and how the management is made by a quality option. (N.C.)

Progress report I.R.S.N. 2003

International Nuclear Information System (INIS)

2003-01-01

This report is divided in six chapters. The first one is devoted to the safety of civil installations, transport and radioactive wastes. In this chapter we find the safety of PWR type reactors, the safety of Cabri reactor, ITER tokamak, the storage of radioactive waste and the criticality of packages with actinides. The second chapter concerns the protection of man and environment, we find the radiation monitoring of workers, the medical management of irradiated personnel, the radiation protection of nuclear power plants. The third chapter is relative to the national defense. The fourth chapter tackles the international relations and the part taken by I.R.S.N. in the Europe construction. The fifth chapter gives the contribution to public information, communication and training. The sixth and last chapter explains the quality policy at I.R.S.N. and how the management is made by a quality option. (N.C.)

The performance of biological anaerobic filters packed with sludge-fly ash ceramic particles (SFCP) and commercial ceramic particles (CCP) during the restart period: effect of the C/N ratios and filter media.

Science.gov (United States)

Yue, Qinyan; Han, Shuxin; Yue, Min; Gao, Baoyu; Li, Qian; Yu, Hui; Zhao, Yaqin; Qi, Yuanfeng

2009-11-01

Two lab-scale upflow biological anaerobic filters (BAF) packed with sludge-fly ash ceramic particles (SFCP) and commercial ceramic particles (CCP) were employed to investigate effects of the C/N ratios and filter media on the BAF performance during the restart period. The results indicated that BAF could be restarted normally after one-month cease. The C/N ratio of 4.0 was the thresholds of nitrate removal and nitrite accumulation. TN removal and phosphate uptake reached the maximum value at the same C/N ratio of 5.5. Ammonia formation was also found and excreted a negative influence on TN removal, especially when higher C/N ratios were applied. Nutrients were mainly degraded within the height of 25 cm from the bottom. In addition, SFCP, as novel filter media manufactured by wastes-dewatered sludge and fly ash, represented a better potential in inhibiting nitrite accumulation, TN removal and phosphate uptake due to their special characteristics in comparison with CCP.

Repair of EL4 leaks

International Nuclear Information System (INIS)

1985-03-01

The reactor shutdown was decided on the 15th of November 1984, because the evolution of the carbon dioxide quantity in the helium blanket of the heavy water. Leaks have been localized on three different channels. Repairs have been made in hard conditions taking into account the reactor state (materials strongly irradiated). The restart has been authorized on the 24th of January 1985 [fr

Renewal of cooling system of JMTR

International Nuclear Information System (INIS)

Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Koike, Sumio; Nishiyama, Yutaka; Fukasaku, Akitomi

2011-06-01

The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. The refurbishment work was started from 2007, and restart is planned in 2011. Renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities whose replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replacement priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

Identifying and Analyzing Novel Epilepsy-Related Genes Using Random Walk with Restart Algorithm

Directory of Open Access Journals (Sweden)

Wei Guo

2017-01-01

Full Text Available As a pathological condition, epilepsy is caused by abnormal neuronal discharge in brain which will temporarily disrupt the cerebral functions. Epilepsy is a chronic disease which occurs in all ages and would seriously affect patients’ personal lives. Thus, it is highly required to develop effective medicines or instruments to treat the disease. Identifying epilepsy-related genes is essential in order to understand and treat the disease because the corresponding proteins encoded by the epilepsy-related genes are candidates of the potential drug targets. In this study, a pioneering computational workflow was proposed to predict novel epilepsy-related genes using the random walk with restart (RWR algorithm. As reported in the literature RWR algorithm often produces a number of false positive genes, and in this study a permutation test and functional association tests were implemented to filter the genes identified by RWR algorithm, which greatly reduce the number of suspected genes and result in only thirty-three novel epilepsy genes. Finally, these novel genes were analyzed based upon some recently published literatures. Our findings implicate that all novel genes were closely related to epilepsy. It is believed that the proposed workflow can also be applied to identify genes related to other diseases and deepen our understanding of the mechanisms of these diseases.

Refurbishment of BR2 (Phases 4 and 5)

Energy Technology Data Exchange (ETDEWEB)

Gubel, P.; Dekeyser, J.; Van Der Auwera, J

1998-07-01

The BR2 is a materials testing reactor and is SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In phase 4 of the refurbishment programme, various activities were performed to allow reactor start-up. In phase 5, remaining refurbishment works were carried out as well as the extra studies and upgradings required by the licensing authorities. Major achievements in 1997 are described and discussed.

Refurbishment of BR2 (Phases 4 and 5)

International Nuclear Information System (INIS)

Gubel, P.; Dekeyser, J.; Van Der Auwera, J.

1998-01-01

The BR2 is a materials testing reactor and is SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In phase 4 of the refurbishment programme, various activities were performed to allow reactor start-up. In phase 5, remaining refurbishment works were carried out as well as the extra studies and upgradings required by the licensing authorities. Major achievements in 1997 are described and discussed

Refurbishment of BR2 (Phases 4 and 5)

Energy Technology Data Exchange (ETDEWEB)

Gubel, P; Dekeyser, J; Van Der Auwera, J

1998-07-01

The BR2 is a materials testing reactor and is SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In phase 4 of the refurbishment programme, various activities were performed to allow reactor start-up. In phase 5, remaining refurbishment works were carried out as well as the extra studies and upgradings required by the licensing authorities. Major achievements in 1997 are described and discussed.

Inter-Disciplinary Collaboration in Support of the Post-Standby TREAT Mission

Energy Technology Data Exchange (ETDEWEB)

DeHart, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Nicolas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jensen, Colby [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parry, James [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, Tony [Idaho State Univ., Pocatello, ID (United States); Phoenix, William [Walsh Engineering Services, Idaho Falls, ID (United States)

2016-09-01

Although analysis methods have advanced significantly in the last two decades, high fidelity multi- physics methods for reactors systems have been under development for only a few years and are not presently mature nor deployed. Furthermore, very few methods provide the ability to simulate rapid transients in three dimensions. Data for validation of advanced time-dependent multi- physics is sparse; at TREAT, historical data were not collected for the purpose of validating three-dimensional methods, let alone multi-physics simulations. Existing data continues to be collected to attempt to simulate the behavior of experiments and calibration transients, but it will be insufficient for the complete validation of analysis methods used for TREAT transient simulations. Hence, a 2018 restart will most likely occur without the direct application of advanced modeling and simulation methods. At present, the current INL modeling and simulation team plans to work with TREAT operations staff in performing reactor simulations with MAMMOTH, in parallel with the software packages currently being used in preparation for core restart (e.g., MCNP5, RELAP5, ABAQUS). The TREAT team has also requested specific measurements to be performed during startup testing, currently scheduled to run from February to August of 2018. These startup measurements will be crucial in validating the new analysis methods in preparation for ultimate application for TREAT operations and experiment design. This document describes the collaboration between modeling and simulation staff and restart, operations, instrumentation and experiment development teams to be able to effectively interact and achieve successful validation work during restart testing.

US/Japan collaborative program on fusion reactor materials: Summary of the tenth DOE/JAERI Annex I technical progress meeting on neutron irradiation effects in first wall and blanket structural materials

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1989-01-01

This meeting was held at Oak Ridge National Laboratory on March 17, 1989, to review the technical progress on the collaborative DOE/JAERI program on fusion reactor materials. The purpose of the program is to determine the effects of neutron irradiation on the mechanical behavior and dimensional stability of US and Japanese austenitic stainless steels. Phase I of the program focused on the effects of high concentrations of helium on the tensile, fatigue, and swelling properties of both US and Japanese alloys. In Phase II of the program, spectral and isotropic tailoring techniques are fully utilized to reproduce the helium:dpa ratio typical of the fusion environment. The Phase II program hinges on a restart of the High Flux Isotope Reactor by mid-1989. Eight target position capsules and two RB* position capsules have been assembled. The target capsule experiments will address issues relating to the performance of austenitic steels at high damage levels including an assessment of the performance of a variety of weld materials. The RB* capsules will provide a unique and important set of data on the behavior of austenitic steels irradiated under conditions which reproduce the damage rate, dose, temperature, and helium generation rate expected in the first wall and blanket structure of the International Thermonuclear Experimental Reactor

Third session: new projects; Session 3: Les Projets nouveaux

Energy Technology Data Exchange (ETDEWEB)

Iracane, D. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France); Dupuy, J.P. [AREVA-Technicatome, 13 - Aix en Provence (France); Perotto, G. [AREVA-Framatome-ANP, 13 - Aix en Provence (France); Ithurralde, G. [Electricite de France (EDF), 13 - Aix en Provence (France); Cabrillat, J.C. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Becle, D. [FRAMATOME, AREVA-FANP, 69 - Lyon (France); Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 75 - Paris (France)

2005-07-01

The RJH (Jules Horowitz reactor) is a 100 MWth research reactor, cooled and moderated by water, that has been designed to operate with high density (8 g U/cm{sup 3}) and low enrichment fuel (19.75%) uranium molybdenum fuel. Its main neutron performances are expected to be: -) maximum fast neutron flux (E > 0.907 MeV) in the sample: 5.10{sup 14} n/cm{sup 2}/s, -) maximum thermal neutron flux (E < 0.625 eV): 4.10{sup 14} n/cm{sup 2}/s, and -) maximal neutron damage: 16 dpa/year. RJH that will be built on the Cea's site of Cadarache, is scheduled to enter into service in 2014, it will take over the activities of the Osiris reactor and its operating life is planned to be over 50 years long. The Cabri research reactor is located on the Cadarache site, it has been operating since 1962 and is now undergoing important upgrading operations whose aim is to make the facility fit to study the behaviour of new nuclear fuels in accidental situations. The RES test reactor is a facility whose aim is to qualify nuclear fuels and core configurations of existing propulsion reactors onboard of aircraft carriers or submarines and to propose innovative concepts for future propulsion reactors. (A.C.)

An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

International Nuclear Information System (INIS)

Behrooz, A.

2008-01-01

The two-phase flow instability is of interest for the design and operation of many industrial systems such as boiling water reactors (BWRs), chemical reactors, and steam generators. In case of BWRs, the flow instabilities are coupled to the power instabilities via neutronic-thermal hydraulic feedbacks. Since these instabilities produce also local pressure oscillations, the coolant flashing plays a very important role at low pressure. Many frequency-domain codes have been used for two-phase flow stability analysis of thermal hydraulic industrial systems with particular emphasis to BWRs. Some were ignoring the effect of the local pressure, or the effect of 3D power oscillations, and many were not able to deal with the neutronics-thermal hydraulics problems considering the entire core and all its fuel assemblies. The new frequency domain tool uses the best available nuclear, thermal hydraulic, algebraic and control theory methods for simulating BWRs and analyzing their stability in either off-line or on-line fashion. The novel code takes all necessary information from plant files via an interface, solves and integrates, for all reactor fuel assemblies divided into a number of segments, the thermal-hydraulic non-homogenous non-equilibrium coupled linear differential equations, and solves the 3D, two-energy-group diffusion equations for the entire core (with spatial expansion of the neutron fluxes in Legendre polynomials).It is important to note that the neutronics equations written in terms of flux harmonics for a discretized system (nodal-modal equations) generate a set of large sparse matrices. The eigenvalue problem associated to the discretized core statics equations is solved by the implementation of the implicit restarted Arnoldi method (IRAM) with implicit shifted QR mechanism. The results of the steady state are then used for the calculation of the local transfer functions and system transfer matrices. The later are large-dense and complex matrices, (their size

Safety test facilities. Needs and concepts. A French evaluation

International Nuclear Information System (INIS)

Tretiakoff, O.; Bailly, J.

1976-01-01

The fuel behavior of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program. These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The purpose of this paper is to explain how a logical set of simple observations has led us to the present state of the Cadarache in-pile experimental safety program and how it may help us to find our way in a dense forest of both technical and psychological difficulties

Annual technical report of the prototype fast breeder reactor Monju. 2011

International Nuclear Information System (INIS)

2012-08-01

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2011. From the aspect of the design evaluation, the following items are summarized: 1) the evaluation of the decay heat removal of Monju core by natural convection, and the safety measures against earthquake and tsunami, which were carried out from the lessons learned at the Fukushima-daiichi accident due to the Great East Japan Earthquake on March 11, 2011, 2) the control rod worth confirmation and the evaluation of nuclear data library based on the data of Core Confirmation Test, which is the first step of Monju system startup test restarted in 2010, 3) the evaluation of the hydrogen concentration behavior, which detects the leak of water from the heat transfer tube of steam generator. Then, from the aspect of the maintenance technology, the following items are summarized: 1) the results of the function confirmation test on the water/steam system, after the long-term suspension, 2) confirmation of the integrity of cracked cylinder liners of emergency diesel generator, 3) replacement of the annulus ventilation duct, 4) evaluation of reduction of the periodic inspection schedule after full power operation. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

Experimental RA reactor operation with 80% enriched fuel - Program of experimental operation: a) Program of experimental operation with 80% enriched fuel at low power, b) contents of the experimental operation with 80% enriched fuel at higher power levels; Program probnog rada: a) Program probnog rada reaktora sa 80% obogacenim gorivom na malim snagama, b) sadrzaj programa probnog rada reaktora RA sa 80% obogacenim gorivom na vecim snagama

Energy Technology Data Exchange (ETDEWEB)

Martinc, R; Sotic, O; Skoric, M; Cupac, S; Bulovic, V; Maric, I; Marinkov, L [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1980-10-15

Highly enriched (80%) uranium oxide fuel was regularly used in the mixed reactor core with the 2% enriched fuel since 1976. The most important changes related to reactor operation, in comparison with the original design project were related to reactor core fuelling schemes. At the end of 1979 reactor was shutdown due to the corrosion coating noticed on some fuel elements and due to decrease quality of the heavy water. Subsequently the Sanitary inspector of Serbia has prohibited further reactor operation. Restart of the reactor will not be a simple continuation of operation. It is indispensable to perform complete experimental program including measurements of critical parameters at different power levels for the core with fresh 80% enriched fuel. The aim of this document is to obtain working permission and its contents are in agreement with the procedure demanded by the Safety Committee of the Institute. It includes results of optimization and safety analysis for the initial reactor core. Since the permission for restart is not obtained, a separate RA reactor safety report is prepared in addition to the program for experimental operation. This report includes: detailed program for reactor experimental operation with 80% enriched fuel in the core at low power levels, and contents of the experimental operation with 80% enriched fuel in the core at higher power levels. [Serbo-Croat] Od decembra 1976. godine redovno je korisceno 80% obogaceno gorivo u mesanoj resetki reaktorskog jezgra sa 2% obogacenim gorivom. Najvece izmene na reaktoru u odnosu na originalni projekat izvrsene su u nacinu rukovanja gorivom. Krajem marta 1979. godine obustavljen je rad reaktora usled naslaga na gorivnim elementima i loseg stanja teske vode. Naknadno je izdata zabrana za rad reaktora od strane Sanitarnog inspektora SR Srbije. Ponovno pustanje reaktora u rad nece biti jednostavan nastavak rada. Neophodno je da se izvede kompletan program merenja kriticnih parametara i drugih

Overview of the one-day meeting on Small Power Reactors

International Nuclear Information System (INIS)

Rogers, J.M.; Butland, A.T.D.

1991-01-01

The purpose of the meeting was to provide an opportunity for various vendors of small power reactors (under 1MW) to present their ideas and designs, and to subsequently respond to comments and criticisms from an informed audience. Two contrasting papers were presented on the economics of small plant. The perceived view within Nuclear Electric of the economic targets for small PWR reactor designs within the United Kingdom was outlined. The designs under consideration are AP600 and SIR. The presentation concentrated on the various factors making up the overall generating cost, in particular the capital charges, stressing the need to account for the design and launch costs and additional risk of the first-of-a-kind plant. The main conclusion was that small plant were likely to be less economic than large, although utilities may be attracted by this option as a method of restarting a nuclear programme, and that the UK targets would be around Pound 1100/kW installed and 3-3.5 p/kW h. The SIR plant was then used to show how these goals could be achievable, taking series-order factors into account, even if the overnight installed cost for a single plant is more expensive compared with a large plant. In another paper the application of passive safety features within designs was considered given the present UK safety licensing framework regarding diversity, redundancy, the single failure criterion, and past precedent. Aspects suggested to require special attention were deliberate depressurization of the primary circuit, secureness of hot shutdown, and the adequacy of natural circulation for containment cooling. The rest of the papers dealt with various aspects of the AP600 and SIR designs. (author)

A new impetus for the E&T on fast neutrons reactors in Europe: Incentives, status, perspectives

International Nuclear Information System (INIS)

Renault, C.; Safieh, J.; Figuet, J.

2013-01-01

Summary and conclusions: • The attractive and challenging scientific topics associated to innovative FNRs create a new incentive context for students and young scientists with high potential to embark on a nuclear career. • The perspective of the construction of demonstration reactors or prototypes of SFR, LFR and GFR appears as a strong driver. • For SFR, an exemplary and precursory approach in France has permitted to preserve the knowledge and know-how gained during five decades of R&D and to be passed down to future generation. • The continuous operation of the sodium school and of the Phenix plant simulator has created a favourable context to restart E&T courses and tools on SFR. • International E&T surveys have strongly underlined the complementary role of skills and competences, in addition to knowledge, for the qualification of nuclear workers. • For this, E&T infrastructures (simulators, experimental facilities,…) are called to play a major role to complement courses. The development of new infrastructures is considered in Europe

Evaluation of I and C architecture alternatives required for the jupiter Icy moons orbiter (JIMO) reactor

International Nuclear Information System (INIS)

Muhlheim, M. D.; Wood, R. T.; Bryan, W. L.; Wilson Jr, T. L.; Holcomb, D. E.; Korsah, K.; Jagadish, U.

2006-01-01

This paper discusses alternative architectural considerations for instrumentation and control (I and C) systems in high-reliability applications to support remote, autonomous, inaccessible nuclear reactors, such as a space nuclear power plant (SNPP) for mission electrical power and space exploration propulsion. This work supported the pre-conceptual design of the reactor control system for the Jupiter Icy Moons Orbiter (JIMO) mission. Long-term continuous operation without intermediate maintenance cycles forces consideration of alternatives to commonly used active, N-multiple redundancy techniques for high-availability systems. Long space missions, where mission duration can exceed the 50% reliability limit of constituent components, can make active, N-multiple redundant systems less reliable than simplex systems. To extend a control system lifetime beyond the 50% reliability limits requires incorporation of passive redundancy of functions. Time-dependent availability requirements must be factored into the use of combinations of active and passive redundancy techniques for different mission phases. Over the course of a 12 to 20-year mission, reactor control, power conversion, and thermal management system components may fail, and the I and C system must react and adjust to accommodate these failures and protect non-failed components to continue the mission. This requires architectural considerations to accommodate partial system failures and to adapt to multiple control schemes according to the state of non-failed components without going through a complete shutdown and restart cycle. Relevant SNPP I and C architecture examples provide insights into real-time fault tolerance and long-term reliability and availability beyond time periods normally associated with terrestrial power reactor I and C systems operating cycles. I and C architectures from aerospace systems provide examples of highly reliable and available control systems associated with short- and long

Knowledge management implementation on the restart of the nuclear power plant construction

International Nuclear Information System (INIS)

Vetere, C.; Eppenstein, M.

2007-01-01

Full text: Restarting the NPP construction after 10 years of inactivity is a process that involves many preliminary tasks associated with the transition period to get the project started again. Implementing a KM program during the preparatory phase motivates the personnel and facilitates the completion of these additional activities. Human Resources Motivation: Manpower is the most critical aspect to consider at the moment of restarting the NPP project. The reduced engineering teams left at the NPP lost their motivation as a result of the absence of project requirements. These groups, which were responsible of key activities in the past, and now assigned to other tasks, must be reinserted to the schedule and functions required by the project management. Moreover, they constitute the core that would transfer knowledge to the future personnel. Therefore, it is a good practice to include these engineering groups from the very beginning of the KM development. It is proved that the participation of these groups in the KM design and definition, in the knowledge map building, in identifying the domains and performing critical knowledge analysis by means of workshops, and in meetings and individual interviews facilitates the reactivation of them. The demands from the Knowledge Management Project create a good atmosphere to stimulate sharing and competences development. Capturing Experts' Knowledge. During the years of inactivity of the plant construction many professionals and specialists that belonged to the original project teams left the organization taking with them their data and information related to the project evolution, and valuable undocumented knowledge. Documented meetings between current and past experts, or through an Experts Consulting Group articulates this tacit knowledge, and provides a source of answers about previous situations, taken decisions and critical issues. Furthermore, implementing a feedback program prevents the risk of knowledge loss due to

[Leak on underground pipings. Corrosion in containment spray systems at Bugey NPP (Preliminary Information)

International Nuclear Information System (INIS)

1996-01-01

During last refuelling shutdown at BUGEY 3, this year on the fourteenth of February, the plant operator discovered a wide corrosion on the reactor vessel head and its equipments. The reactor head vessel had recently been replaced by a new one since last reactor shutdown in order to treat the vessel head adaptor safety problem. The cause of this corrosion is a small primary leak on this pipe flange. The leak had been found fortuitously during a field inspection of valves while there was not reactor charge, seven months before the reactor was shutdown for refuelling. At this time the primary leak had been leaktighted by closure of a manual valve and the reactor was restarted up

Solid State Division progress report for period ending September 30, 1990

International Nuclear Information System (INIS)

Green, P.H.; Hinton, L.W.

1991-03-01

This report covers research progress in the Solid State Division from April 1, 1989, to September 30, 1990. During this period, division research programs were significantly enhanced by the restart of the High-Flux Isotope Reactor (HFIR) and by new initiatives in processing and characterization of materials

Solid State Division progress report for period ending September 30, 1990

Energy Technology Data Exchange (ETDEWEB)

Green, P.H.; Hinton, L.W. (eds.)

1991-03-01

This report covers research progress in the Solid State Division from April 1, 1989, to September 30, 1990. During this period, division research programs were significantly enhanced by the restart of the High-Flux Isotope Reactor (HFIR) and by new initiatives in processing and characterization of materials.

INES application in Kazakhstan, annual review

International Nuclear Information System (INIS)

Krechetov, S.

1997-01-01

For the period from the last TCM meeting all Kazakhstani facilities operated rather good. Only one event which attracted public attention occurred at BN-350 on 23 March and was rated as level 0. It had been reported according to INES procedure. I would like to suggest the TCM participants some additional information related to our nuclear activities in Kazakhstan. Our first own law of the Atomic energy use put into force on 14th of April 1997. The research reactor WWER-K nearby to Almaty after 8 years under shutdown reached first critically on 18 April under restarting program. The program to restart of the reactor have to be completed at the end of this year. For the improvement of the relations with the public since this year IAEA started to provide our government TV channel the information about safety status and radiation situation at and around of the nuclear facilities on weekly basis

Emergency recirculation pump driving mechanism

International Nuclear Information System (INIS)

Morooka, Shin-ichi.

1980-01-01

Purpose: To sufficiently secure the coolant flow rate in a reactor core and restrict the temperature on the surface of fuel elements to low degree when the coolant is lost in a BWR type reactor. Constitution: In order to secure sufficient coolant flow rate in a reactor core and to sufficiently cool the reactor core when the coolant is lost in a BWR type reactor, it is tripped upon loss of power supply simultaneously when an accident occurs, a recycling pump at the side of normal reactor where its rotating speed is decelerated in accordance with its inertia is restarted by the pressure water stored in a tank out of the reactor to increase the coolant flow rate in the reactor core so as to sufficiently cool the reactor core. (Aizawa, K.)

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

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

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

International conference on fast reactors and related fuel cycles (FR09): Challenges and opportunities. CN-176 presentations

International Nuclear Information System (INIS)

2009-01-01

Renewed interest in nuclear energy is driven by the need to develop carbon free energy sources, by demographics and development in emerging economies, as well as by security of supply concerns. It is expected that nuclear energy will deliver huge amounts of energy to both emerging and developed economies. However, acceptance of large scale contributions would depend on satisfaction of key drivers to enhance sustainability in terms of economics, safety, adequacy of natural resources, waste reduction, non-proliferation and public acceptance. Fast spectrum reactors with recycle enhance the sustainability indices significantly. This has led to the focus on fast spectrum reactors with recycle in the Generation IV International Forum (GIF) and the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative of the IAEA. It is expected that 2009 will register major events in the domain of fast spectrum reactors, that is, the restart of Monju in Japan, the first criticality of the China Experimental Fast Reactor in China, as well as new insights through end-of-life studies in Phenix, France. New fast reactors are expected to be commissioned in the near future: the 500 MW(e) Prototype Fast Breeder Reactor in India and the BN-800 unit in the Russian Federation. Moreover, China, France, India, Japan, Republic of Korea and the United States of America are preparing advanced prototypes/ demonstrations and/or commercial reactors for the 2020-2030 horizon. The necessary condition for successful fast reactor deployment in the near and mid-term is the understanding and assessment of innovative technological and design options, based on both past knowledge and experience, as well as on ongoing research and technology development efforts. In this respect, the need for in-depth international information exchange is underscored by the fact that the last large international fast reactor conference was held as far back as 1991. Since then, progress in research

Terminal-shock and restart control of a Mach 2.5, axisymmetric, mixed compression inlet with 40 percent internal contraction. [wind tunnel tests

Science.gov (United States)

Baumbick, R. J.

1974-01-01

Results of experimental tests conducted on a supersonic, mixed-compression, axisymmetric inlet are presented. The inlet is designed for operation at Mach 2.5 with a turbofan engine (TF-30). The inlet was coupled to either a choked orifice plate or a long duct which had a variable-area choked exit plug. Closed-loop frequency responses of selected diffuser static pressures used in the terminal-shock control system are presented. Results are shown for Mach 2.5 conditions with the inlet coupled to either the choked orifice plate or the long duct. Inlet unstart-restart traces are also presented. High-response inlet bypass doors were used to generate an internal disturbance and also to achieve terminal-shock control.

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

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

1989-01-01

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

Safety test facilities. Needs and concepts. A French evaluation

International Nuclear Information System (INIS)

Tretiakoff, O.; Bailly, J.

1976-01-01

The fuel behaviour of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program (local blockage, sudden flow reduction and pump coast-down at constant power). These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The paper explains how a logical set of simple observations has led to the present state of the Cadarache in-pile experimental safety program and how it may help to find the way in a dense forest of both technical and psychological difficulties

SPARC: Demonstrate burst-buffer-based checkpoint/restart on ATS-1.

Energy Technology Data Exchange (ETDEWEB)

Oldfield, Ron A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ulmer, Craig D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Widener, Patrick [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ward, H. Lee [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2018-01-01

Recent high-performance computing (HPC) platforms such as the Trinity Advanced Technology System (ATS-1) feature burst buffer resources that can have a dramatic impact on an application’s I/O performance. While these non-volatile memory (NVM) resources provide a new tier in the storage hierarchy, developers must find the right way to incorporate the technology into their applications in order to reap the benefits. Similar to other laboratories, Sandia is actively investigating ways in which these resources can be incorporated into our existing libraries and workflows without burdening our application developers with excessive, platform-specific details. This FY18Q1 milestone summaries our progress in adapting the Sandia Parallel Aerodynamics and Reentry Code (SPARC) in Sandia’s ATDM program to leverage Trinity’s burst buffers for checkpoint/restart operations. We investigated four different approaches with varying tradeoffs in this work: (1) simply updating job script to use stage-in/stage out burst buffer directives, (2) modifying SPARC to use LANL’s hierarchical I/O (HIO) library to store/retrieve checkpoints, (3) updating Sandia’s IOSS library to incorporate the burst buffer in all meshing I/O operations, and (4) modifying SPARC to use our Kelpie distributed memory library to store/retrieve checkpoints. Team members were successful in generating initial implementation for all four approaches, but were unable to obtain performance numbers in time for this report (reasons: initial problem sizes were not large enough to stress I/O, and SPARC refactor will require changes to our code). When we presented our work to the SPARC team, they expressed the most interest in the second and third approaches. The HIO work was favored because it is lightweight, unobtrusive, and should be portable to ATS-2. The IOSS work is seen as a long-term solution, and is favored because all I/O work (including checkpoints) can be deferred to a single library.

Romanian nuclear fuel fabrication and in-reactor fuel operational experience

International Nuclear Information System (INIS)

Budan, O.

2003-01-01

A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

Energy Technology Data Exchange (ETDEWEB)

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

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

Return to Chernobyl

Energy Technology Data Exchange (ETDEWEB)

Nosovsky, Anatolij

1995-09-01

Despite the catastrophic accident at the Chernobylsk 4 reactor in 1986, the Ukraine is currently expanding its nuclear industry. The government is committed to increasing the share of nuclear output to 40% of the country`s electric power and the Chernobyl plant is included in this plan. All the Chernobyl reactors were closed down at the time of the accident, but units 1, 2 and 3 had all been restarted after safety modifications by December 1987. A fire in the turbine hall of unit 2 in 1991 resulted in the closure of that reactor and precipitated a political decision to close the entire plant by 1993. The economic consequences of such action and the safe operation of the remaining two reactors led, however, to the reversal of that decision. Work is now far advanced on unit 2 for a restart in 1996 and the management wants to upgrade all three reactors according to IAEA guidelines. Nevertheless, the question of closure of the Chernobyl plant remains in the air. A conditional acceptance of closure by 2000 has been made by the Ukraine provided the shortfall in power is taken up by a new gas-fired station. International finance is being sought for decommissioning, for urgent action on the decaying sarcophagus of unit 4, and for the gas-fired plant. Closure of the plant, given the social upheaval of the accident and recent political events, could contribute to the health of the Ukrainian national psyche. (UK).

Return to Chernobyl

International Nuclear Information System (INIS)

Nosovsky, Anatolij.

1995-01-01

Despite the catastrophic accident at the Chernobylsk 4 reactor in 1986, the Ukraine is currently expanding its nuclear industry. The government is committed to increasing the share of nuclear output to 40% of the country's electric power and the Chernobyl plant is included in this plan. All the Chernobyl reactors were closed down at the time of the accident, but units 1, 2 and 3 had all been restarted after safety modifications by December 1987. A fire in the turbine hall of unit 2 in 1991 resulted in the closure of that reactor and precipitated a political decision to close the entire plant by 1993. The economic consequences of such action and the safe operation of the remaining two reactors led, however, to the reversal of that decision. Work is now far advanced on unit 2 for a restart in 1996 and the management wants to upgrade all three reactors according to IAEA guidelines. Nevertheless, the question of closure of the Chernobyl plant remains in the air. A conditional acceptance of closure by 2000 has been made by the Ukraine provided the shortfall in power is taken up by a new gas-fired station. International finance is being sought for decommissioning, for urgent action on the decaying sarcophagus of unit 4, and for the gas-fired plant. Closure of the plant, given the social upheaval of the accident and recent political events, could contribute to the health of the Ukrainian national psyche. (UK)

DNA replication restart and cellular dynamics of Hef helicase/nuclease protein in Haloferax volcanii.

Science.gov (United States)

Lestini, Roxane; Delpech, Floriane; Myllykallio, Hannu

2015-11-01

Understanding how frequently spontaneous replication arrests occur and how archaea deal with these arrests are very interesting and challenging research topics. Here we will described how genetic and imaging studies have revealed the central role of the archaeal helicase/nuclease Hef belonging to the XPF/MUS81/FANCM family of endonucleases in repair of arrested replication forks. Special focus will be on description of a recently developed combination of genetic and imaging tools to study the dynamic localization of a functional Hef::GFP (Green Fluorescent Protein) fusion protein in the living cells of halophilic archaea Haloferax volcanii. As Archaea provide an excellent and unique model for understanding how DNA replication is regulated to allow replication of a circular DNA molecule either from single or multiple replication origins, we will also summarize recent studies that have revealed peculiar features regarding DNA replication, particularly in halophilic archaea. We strongly believe that fundamental knowledge of our on-going studies will shed light on the evolutionary history of the DNA replication machinery and will help to establish general rules concerning replication restart and the key role of recombination proteins not only in bacteria, yeast and higher eukaryotes but also in archaea. Copyright © 2015 Elsevier B.V. and Société Française de Biochimie et Biologie Moléculaire (SFBBM). All rights reserved.

Fast parallel MR image reconstruction via B1-based, adaptive restart, iterative soft thresholding algorithms (BARISTA).

Science.gov (United States)

Muckley, Matthew J; Noll, Douglas C; Fessler, Jeffrey A

2015-02-01

Sparsity-promoting regularization is useful for combining compressed sensing assumptions with parallel MRI for reducing scan time while preserving image quality. Variable splitting algorithms are the current state-of-the-art algorithms for SENSE-type MR image reconstruction with sparsity-promoting regularization. These methods are very general and have been observed to work with almost any regularizer; however, the tuning of associated convergence parameters is a commonly-cited hindrance in their adoption. Conversely, majorize-minimize algorithms based on a single Lipschitz constant have been observed to be slow in shift-variant applications such as SENSE-type MR image reconstruction since the associated Lipschitz constants are loose bounds for the shift-variant behavior. This paper bridges the gap between the Lipschitz constant and the shift-variant aspects of SENSE-type MR imaging by introducing majorizing matrices in the range of the regularizer matrix. The proposed majorize-minimize methods (called BARISTA) converge faster than state-of-the-art variable splitting algorithms when combined with momentum acceleration and adaptive momentum restarting. Furthermore, the tuning parameters associated with the proposed methods are unitless convergence tolerances that are easier to choose than the constraint penalty parameters required by variable splitting algorithms.

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

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

New tensioning equipment saves time at Browns Ferry

International Nuclear Information System (INIS)

Wagner, P.

1996-01-01

Improved reactor vessel tensioning technology has proved very successful at TVA's Browns Ferry station. The new equipment was used during unit 2's autumn 1994 outage and will be used there at the next outage, scheduled for March 1996. It will be deployed at unit 3, currently in restart mode. (author)

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

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

The roles of the seismic safety and monitoring systems in the PEC fast reactor

International Nuclear Information System (INIS)

Masoni, P.; Di Tullio, E.M.; Massa, B.; Martelli, A.; Sano, T.

1988-01-01

Two different seismic systems are foreseen in the case of PEC: the seismic safety system, that provides the automatic scram, and the seismic monitoring system. During earthquake, three triaxial seismic switches are triggered if a threshold value of the ground acceleration is exceeded. In this case, the signals from the seismic switches are processed by the safety system (with a 2/3 logic) and the shutdown system is triggered. Peak acceleration is the parameter used by the safety system to quantify the seismic event. This way, however, no information is obtained with regard to earthquake frequency content. Thus, reactor safety is guaranteed by adopting a threshold considerably lower than the Z.P.A. of the Design Basis Earthquake. Furthermore, in the case of significant earthquakes, the seismic motion is measured by about 20 triaxial accelerometers, located both in the free field and on the plant's structures. Data are digitazed and recordered by the seismic monitoring system. This system also elaborates the recordered time-histories providing floor response spectra and compares such spectra to the design values. The above-mentioned elaborations and comparisons are performed in short time for two triaxial measuring positions, thus allowing the Operator to immediately get a more complete information on the seismic event. The complete set of data recorded by the seismic monitoring system also allows the actual dynamic response of the plant to be determined and compared to the design values. On the basis of this comparison the necessary safety analysis can be carried out to verify whether the design limits of the plant were respected: in the positive case the reactor can be restarted. (author)

Safety evaluation for the prototype Fast Breeder Reactor MONJU as a Japanese TSO

International Nuclear Information System (INIS)

Endo, Hiroshi

2010-01-01

In the safety field of fast breeder reactors (FBRs), JNES is conducting an evaluation work of the safety regulation by Nuclear and Industry Safety Agency (NISA) for the re-start of a prototype FBR MONJU. MONJU has been stopped over 14 years since 1995 due to a sodium leakage accident at a secondary heat transport system, and is now reached to the criticality on 8th of May, 2010. JNES is supporting the safety regulation work conducted by NISA based on the following activities: i) Support of the technical evaluation of the application for the establishment license prepared by Japan Atomic Energy Agency (JAEA), ii) Support of the description of the safety review report by NISA based on independent safety analyses for the major accident events such as unprotected loss-of-flow (ULOF) by employing the latest findings on the study of core disruptive accidents (CDAs) independently conducted by JNES, iii) Support of the risk-informed-regulation (RIR) such as an accident management (AM) review, iv), and Consideration on the safety regulation policy from the points of severe accidents and source-term behaviors including the cesium (Cs). The objective of this paper is to introduce the major activities of JNES in the safety domain of MONJU regulations. (author)

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

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

1987-01-01

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

Nuclear situation in Japan

International Nuclear Information System (INIS)

2006-01-01

This analysis takes stock on the nuclear situation in Japan. It discusses the ambitious equipment program in collaboration with the France, the destabilization of the japanese nuclear industry following the accidents and the energy policy evolutions. It presents the projects of the japanese nuclear industry: the Monju reactor restart, the Pluthermal project, the reprocessing power plant of Rokkasho Mura, the new reactors, the russian weapons dismantling, the ITER site selection and the buy out of Westinghouse by Toshiba. (A.L.B.)

Nuclear situation in Japan; La situation du nucleaire au Japon

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

This analysis takes stock on the nuclear situation in Japan. It discusses the ambitious equipment program in collaboration with the France, the destabilization of the japanese nuclear industry following the accidents and the energy policy evolutions. It presents the projects of the japanese nuclear industry: the Monju reactor restart, the Pluthermal project, the reprocessing power plant of Rokkasho Mura, the new reactors, the russian weapons dismantling, the ITER site selection and the buy out of Westinghouse by Toshiba. (A.L.B.)

Operation of the BR2 Reactor

International Nuclear Information System (INIS)

Gubel, P.

2006-01-01

The BR2 is still SCK-CEN's most important nuclear facility. After an extensive refurbishment of 22 months to compensate for the ageing of the installations, to enhance the reliability of operation and to comply with modern safety standards, it was restarted in April 1997. The facility is mainly used for the irradiation and testing of fuels and materials and for commercial productions - including radioisotopes for the medical and industrial uses, and NTD-Silicon. The article describes the main achievements and activities in 2005

Operation of the BR2 Reactor

International Nuclear Information System (INIS)

Gubel, P.

2005-01-01

The BR2 is still SCK-CEN's most important nuclear facility. After an extensive refurbishment of 22 months to compensate for the ageing of the installations, to enhance the reliability of operation and to comply with modern safety standards, it was restarted in April 1997. The facility is mainly used for the irradiation and testing of fuels and materials and for commercial productions - including radioisotopes for the medical and industrial uses, and NTD-Silicon. The article describes the main activities and achievements in 2004

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

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

Modelling of the thermomechanical and physical processes in FR fuel pins using the GERMINAL code

International Nuclear Information System (INIS)

Roche, L.; Pelletier, M.

2000-01-01

In the frame of the R and D on Fast Reactor mixed oxide fuels, CEA/DEC has developed the computer code GERMINAL for studying fuel pin thermal and mechanical behaviour, both during steady-state and incidental conditions, up to high burn-up (25 at%). The first part of this paper is devoted to the description of the main models: fuel evolution (central hole and porosity evolution, Plutonium redistribution, O/M radial profile, transient gas swelling, melting fuel behaviour, minor actinides production), high burn-up models (fission gas, volatile fission products and JOG formation), fuel-cladding heat transfer, fuel-cladding mechanical interaction. The second part gives some examples of calculation results taken from the GERMINAL validation data base (more than 40 experiments from PHENIX, PFR, CABRI reactors), with special emphasis on: local fission gas retention and global release, fuel geometry evolution, radial redistribution of plutonium for high burn-up fuels, solid and annular fuel behaviour during power ramps including fuel melting, helium formation from MA (Am and Np) doped homogeneous fuels. (author)

Investigation on shortening fabrication process of instrumented irradiation capsule of JMTR

International Nuclear Information System (INIS)

Nagata, Hiroshi; Inoue, Shuichi; Yamaura, Takayuki; Tsuchiya, Kunihiko; Nagao, Yoshiharu

2013-06-01

Refurbishment of The Japan Materials Testing Reactor (JMTR) was completed in FY2010. For damage caused by the 2011 off the Pacific coast of Tohoku Earthquake, the repair of facilities was completed in October 2012. Currently, the JMTR is in preparation for restart. Irradiation tests for LWRs safety research, science and technologies and production of RI for medical diagnosis medicine, etc. are expected after the JMTR restart. On the other hand, aiming at the attractive irradiation testing reactor, the usability improvement has been discussed. As a part of the usability improvement, shortening of turnaround time to get irradiation results from an application for irradiation use was discussed focusing on the fabrication process of irradiation capsules, where the fabrication process was analyzed and reviewed by referring a trial fabrication of the mockup capsule. As a result, it was found that the turnaround time can be shortened 2 months from fabrication period of 6 months with communize of irradiation capsule parts, application of ready-made instrumentation including the sheath heater, reconsideration of inspection process, etc. (author)

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

A review of the UK fast reactor programme, March 1981

International Nuclear Information System (INIS)

Smith, R.D.

1981-01-01

steam cycle, a re-evaluation of various boiler design styles have been undertaken. The types being studied cover straight tube, U and J tube, hockey stick and helical coil designs. A final choice of steam generator for CDFR has not yet been made. Key features in design selection are overall integrity to ensure high availability, ease of maintenance and repair following minor tube leaks, resistance to escalation of minor leaks to more major sodium/water reaction incidents. inspectability, relevance of previous experience and capital cost. Operation of the PFR was interrupted by a series of small leaks in the tube-to-tube plate welds in the three evaporator units. The key areas have now been shot-peened to remove tensile stresses in the surface material. the units have been repaired and operation at high power should be restarted during March 1981. The performance of the core and primary circuits has continued to be excellent. A major step forward was taken when the first 16 sub-assemblies were reprocessed in the PFR reprocessing plant. During the year a system economic study has been completed to assist in providing a sound technical basis for a fast reactor inquiry. This study was based on a set of programme assumptions and economic ground rules developed and agreed in 1978 within the UK nuclear industry and updated in 1979. Broadly, it was concluded that the introduction of the fast reactor as soon and as rapidly as feasible also leads, even on conservative uranium price assumptions. to economic benefits, the benefits being greatest when electrical demand and raw fuel prices are at their highest. The benefits are little affected by 25% increases in costs, except for an increase in fast reactor capital cost when uranium stays cheap; in this case the benefit virtually disappears if it is assumed that all other capital costs stay constant. Use of the fast reactor also produces a dramatic fall (almost three-fold) in the need uranium imports and, if market forces are

A review of the UK fast reactor programme, March 1981

Energy Technology Data Exchange (ETDEWEB)

Smith, R D [Risley Nuclear Power Development Establishment, Risley, Warrington (United Kingdom)

1981-05-01

steam cycle, a re-evaluation of various boiler design styles have been undertaken. The types being studied cover straight tube, U and J tube, hockey stick and helical coil designs. A final choice of steam generator for CDFR has not yet been made. Key features in design selection are overall integrity to ensure high availability, ease of maintenance and repair following minor tube leaks, resistance to escalation of minor leaks to more major sodium/water reaction incidents. inspectability, relevance of previous experience and capital cost. Operation of the PFR was interrupted by a series of small leaks in the tube-to-tube plate welds in the three evaporator units. The key areas have now been shot-peened to remove tensile stresses in the surface material. the units have been repaired and operation at high power should be restarted during March 1981. The performance of the core and primary circuits has continued to be excellent. A major step forward was taken when the first 16 sub-assemblies were reprocessed in the PFR reprocessing plant. During the year a system economic study has been completed to assist in providing a sound technical basis for a fast reactor inquiry. This study was based on a set of programme assumptions and economic ground rules developed and agreed in 1978 within the UK nuclear industry and updated in 1979. Broadly, it was concluded that the introduction of the fast reactor as soon and as rapidly as feasible also leads, even on conservative uranium price assumptions. to economic benefits, the benefits being greatest when electrical demand and raw fuel prices are at their highest. The benefits are little affected by 25% increases in costs, except for an increase in fast reactor capital cost when uranium stays cheap; in this case the benefit virtually disappears if it is assumed that all other capital costs stay constant. Use of the fast reactor also produces a dramatic fall (almost three-fold) in the need uranium imports and, if market forces are

Identifying novel fruit-related genes in Arabidopsis thaliana based on the random walk with restart algorithm.

Science.gov (United States)

Zhang, Yunhua; Dai, Li; Liu, Ying; Zhang, YuHang; Wang, ShaoPeng

2017-01-01

Fruit is essential for plant reproduction and is responsible for protection and dispersal of seeds. The development and maturation of fruit is tightly regulated by numerous genetic factors that respond to environmental and internal stimulation. In this study, we attempted to identify novel fruit-related genes in a model organism, Arabidopsis thaliana, using a computational method. Based on validated fruit-related genes, the random walk with restart (RWR) algorithm was applied on a protein-protein interaction (PPI) network using these genes as seeds. The identified genes with high probabilities were filtered by the permutation test and linkage tests. In the permutation test, the genes that were selected due to the structure of the PPI network were discarded. In the linkage tests, the importance of each candidate gene was measured from two aspects: (1) its functional associations with validated genes and (2) its similarity with validated genes on gene ontology (GO) terms and KEGG pathways. Finally, 255 inferred genes were obtained, subsequent extensive analysis of important genes revealed that they mainly contribute to ubiquitination (UBQ9, UBQ8, UBQ11, UBQ10), serine hydroxymethyl transfer (SHM7, SHM5, SHM6) or glycol-metabolism (HXKL2_ARATH, CSY5, GAPCP1), suggesting essential roles during the development and maturation of fruit in Arabidopsis thaliana.

Chernobyl - ''Soviet radiation is the best in the world''

International Nuclear Information System (INIS)

1995-01-01

The background to the negotiations between Ukraine and the G7 countries over the final shutdown of the remaining three Chernobyl reactors is examined. The protracted nature of the negotiations, their complexity and lack of progress reflect the technological, political and socio-economic ambiguities surrounding the plant. The issues involved concern the safety of RMBK reactors, the need in the Ukraine for the power supplied by the Chernobyl units and the requirement to find a permanent solution to the stability of the Sarcophagus. Perceptions of the safety of the remaining reactors differ and closure of the plant on reactor safety grounds would raise questions about the other Russian RMBK reactors. Neither Ukraine nor Russia could afford to replace their RMBKs and cannot do without the power they generate. After the closure in the 1980s, the Chernobyl plant would not have restarted had the West provided a 2000MW gas-fired replacement. But having restarted, Ukraine would not close the plant immediately even if financial aid now became available because of the hardship its loss would inflict on the population while a substitute plant was built. Such financial help as is being offered by the West is not in the form of a free gift but has taken the form of a commercial deal which benefits plant manufacturers in Europe and the USA and which Ukraine views with suspicion. Help is certainly needed for the Sarcophagus, the failure of which would have far-reaching political consequences for the global nuclear industry. (UK)

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

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

International Nuclear Information System (INIS)

Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

2012-01-01

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

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

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983; Istrazivacki nuklearni reaktor RA - Deo I - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1983. godini

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Raickovic, N; Radivojevic, J; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1983-12-15

After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW. [Serbo-Croat] Pregledom nuklearnog goriva iz tehnoloskih kanala reaktora RA koji je izvrsen u periodu decembear 1982-feburuar 1983. godine nakon zaustavljanja reaktora po isteku novembarske kampanje 1982. godine, ustanovljeno je da ponovo dolazi do stvaranja taloga u obliku hidratisanih oksida aluminiuma na kosuljicama gorivnih elemenata. Nakon ponovnog pustanja u rad, reaktor je do novembra 1981. godine neprekidno bio u pogonu na snagama 1,8 - 2 MW. Jezgro je bilo formirano iskljucivo sa od gorivnih elemenata sa 80% obogacenim uran dioksidom. Utvrdjeno je da kosuljica gorivnog elementa nije korodirala, i da se bez nastval rada ne moze utvrditi uzrok pojave taloga na povrsini kosuljice. Da bi se mogucnost korozije aluminjumskih komponenti u primarnom kolu raktora svela na sto manju meru odluceno je da se vrsi neprekidno preciscavanje teske vode i da se istovremeno poveca protok teske vode radom dve pumpe, Komitet za sigurnost Instituta odobrio je ovakav nacin

Experience of boron neutron capture therapy in Japan

International Nuclear Information System (INIS)

Kanda, K.

2004-01-01

Four research reactors are currently licensed for medical application in Japan. As of July 1995, approximately 210 clinical irradiations using these research reactors have been done for brain and skin tumors as shown. The number of chief medical doctors certified by the Government is eleven so far. Among them, eight doctors have already treated tumor patients using the Kyoto University Reactor (KUR, 5MW). Recently in USA clinical trials have been restarted using epithermal neutrons at MIT and BNL. In this paper, the experience of clinical trials of boron neutron capture therapy (BNCT) which have been performed in Japan, mainly physics studies, are reviewed, and current studies are also introduced

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

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

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

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

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

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

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

RIA Fuel Codes Benchmark - Volume 1

International Nuclear Information System (INIS)

Marchand, Olivier; Georgenthum, Vincent; Petit, Marc; Udagawa, Yutaka; Nagase, Fumihisa; Sugiyama, Tomoyuki; Arffman, Asko; Cherubini, Marco; Dostal, Martin; Klouzal, Jan; Geelhood, Kenneth; Gorzel, Andreas; Holt, Lars; Jernkvist, Lars Olof; Khvostov, Grigori; Maertens, Dietmar; Spykman, Gerold; Nakajima, Tetsuo; Nechaeva, Olga; Panka, Istvan; Rey Gayo, Jose M.; Sagrado Garcia, Inmaculada C.; Shin, An-Dong; Sonnenburg, Heinz Guenther; Umidova, Zeynab; Zhang, Jinzhao; Voglewede, John

2013-01-01

Reactivity-initiated accident (RIA) fuel rod codes have been developed for a significant period of time and they all have shown their ability to reproduce some experimental results with a certain degree of adequacy. However, they sometimes rely on different specific modelling assumptions the influence of which on the final results of the calculations is difficult to evaluate. The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burnup and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including the review of experimental approaches as well as the interpretation and use of experimental data relevant for safety. As a contribution to this task, WGFS conducted a RIA code benchmark based on RIA tests performed in the Nuclear Safety Research Reactor in Tokai, Japan and tests performed or planned in CABRI reactor in Cadarache, France. Emphasis was on assessment of different modelling options for RIA fuel rod codes in terms of reproducing experimental results as well as extrapolating to typical reactor conditions. This report provides a summary of the results of this task. (authors)

L-Lake fish: L-Lake/Steel Creek Biological Monitoring Program, January 1986--December 1991

International Nuclear Information System (INIS)

Sayers, R.E. Jr.; Mealing, H.G. III

1992-04-01

The L Lake Biological Monitoring Program was designed to meet environmental regulatory requirements associated with the re-start of L-Reactor and address portions of Section 316(a) of the Clean Water Act, which requires an applicant for a discharge permit to provide scientific evidence that the discharge causes no significant impact on the indigenous ecosystem. The Department of Energy (DOE) must demonstrate that the discharge of L-Reactor effluent into L Lake will not inhibit the eventual establishment of a ''Balanced Biological Community'' (BBC) in at least 50% of the lake

Configuration Management Program Plan

International Nuclear Information System (INIS)

1991-01-01

Westinghouse Savannah River Company (WSRC) has established a configuration management (CM) plan to execute the SRS CM Policy and the requirements of the DOE Order 4700.1. The Reactor Restart Division (RRD) has developed its CM Plan under the SRS CM Program and is implementing it via the RRD CM Program Plan and the Integrated Action Plan. The purpose of the RRD CM program is to improve those processes which are essential to the safe and efficient operation of SRS production reactors. This document provides details of this plan

L-Lake fish: L-Lake/Steel Creek Biological Monitoring Program, January 1986--December 1991

Energy Technology Data Exchange (ETDEWEB)

Sayers, R.E. Jr.; Mealing, H.G. III [Normandeau Associates, Inc., New Ellenton, SC (United States)

1992-04-01

The L Lake Biological Monitoring Program was designed to meet environmental regulatory requirements associated with the re-start of L-Reactor and address portions of Section 316(a) of the Clean Water Act, which requires an applicant for a discharge permit to provide scientific evidence that the discharge causes no significant impact on the indigenous ecosystem. The Department of Energy (DOE) must demonstrate that the discharge of L-Reactor effluent into L Lake will not inhibit the eventual establishment of a ``Balanced Biological Community`` (BBC) in at least 50% of the lake.

Alternatives Analysis for the Resumption of Transient Testing Program

Energy Technology Data Exchange (ETDEWEB)

Lee Nelson

2013-11-01

An alternatives analysis was performed for resumption of transient testing. The analysis considered eleven alternatives – including both US international facilities. A screening process was used to identify two viable alternatives from the original eleven. In addition, the alternatives analysis includes a no action alternative as required by the National Environmental Policy Act (NEPA). The alternatives considered in this analysis included: 1. Restart the Transient Reactor Test Facility (TREAT) 2. Modify the Annular Core Research Reactor (ACRR) which includes construction of a new hot cell and installation of a new hodoscope. 3. No Action

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

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

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

International Nuclear Information System (INIS)

Amano, Yukiya

2009-01-01

Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Planktonic foraminiferal biostratigraphy of the Upper Barremian and Aptian of Crimea

Science.gov (United States)

Brovina, E. A.

2017-09-01

Reexamination of the Barremian-Aptian planktonic foraminifers from three sections (Verkhoirechie, mountain Krasnaya, and Marino) allowed the biostratigraphic scheme for Southwest and Central Crimea to be refined and updated. The following standard zones are recognized in the studied sections: Blowiella blowi (upper Barremian), Hedbergella excelsa (upper Barremian-lower Aptian), Leupoldina cabri (lower Aptian), H. luterbacheri, Globigerinelloides ferreolensis, Gl. barri, Gl. algerianus, Hedbergella trocoidea, Paraticinella rohri (upper Aptian). Beds with Hedbergella ruka are recognized in the B. blowi Zone. Foraminifers from the Partizanskoe section, representing the lower Aptian L. cabri and H. luterbacheri zones, are studied. The recognized strata are correlated with ammonite and nannoplankton zones and paleomagnetic data.

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

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

2000-01-01

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

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Restarting Anticoagulant Treatment After Intracranial Hemorrhage in Patients With Atrial Fibrillation and the Impact on Recurrent Stroke, Mortality, and Bleeding: A Nationwide Cohort Study.

Science.gov (United States)

Nielsen, Peter Brønnum; Larsen, Torben Bjerregaard; Skjøth, Flemming; Gorst-Rasmussen, Anders; Rasmussen, Lars Hvilsted; Lip, Gregory Y H

2015-08-11

Intracranial hemorrhage is the most feared complication of oral anticoagulant treatment. The optimal treatment option for patients with atrial fibrillation who survive an intracranial hemorrhage remains unknown. We hypothesized that restarting oral anticoagulant treatment was associated with a lower risk of stroke and mortality in comparison with not restarting. Linkage of 3 Danish nationwide registries in the period between 1997 and 2013 identified patients with atrial fibrillation on oral anticoagulant treatment with incident intracranial hemorrhage. Patients were stratified by treatment regimens (no treatment, oral anticoagulant treatment, or antiplatelet therapy) after the intracranial hemorrhage. Event rates were assessed 6 weeks after hospital discharge and compared with Cox proportional hazard models. In 1752 patients (1 year of follow-up), the rate of ischemic stroke/systemic embolism and all-cause mortality (per 100 person-years) for patients treated with oral anticoagulants was 13.6, in comparison with 27.3 for nontreated patients and 25.7 for patients receiving antiplatelet therapy. The rate of ischemic stroke/systemic embolism and all-cause mortality (per 100 person-years) for recurrent intracranial hemorrhage, the rate of ischemic stroke/systemic embolism, and all-cause mortality (per 100 person-years) patients treated with oral anticoagulants was 8.0, in comparison with 8.6 for nontreated patients and 5.3 for patients receiving antiplatelet therapy. The adjusted hazard ratio of ischemic stroke/systemic embolism and all-cause mortality was 0.55 (95% confidence interval, 0.39-0.78) in patients on oral anticoagulant treatment in comparison with no treatment. For ischemic stroke/systemic embolism and for all-cause mortality, hazard ratios were 0.59 (95% confidence interval, 0.33-1.03) and 0.55 (95% confidence interval, 0.37-0.82), respectively. Oral anticoagulant treatment was associated with a significant reduction in ischemic stroke/all-cause mortality

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

Model development of SAS4A and investigation on the initiating phase consequences in LMFRs related with material motion

International Nuclear Information System (INIS)

Niwa, H.

1994-01-01

This paper focuses on an analytical aspect of the initiating phase scenario and consequences of postulated core disruptive accident in liquid-metal-cooled fast breeder reactors. An analytical code, SAS4A, has been developed at Argonne National Laboratory, and introduced to PNC. Improvement and validation effort have been performed for the mixed-oxide version of SAS4A at PNC. This paper describes firstly recent development of SAS4A's material motion related models briefly. A fission gas mass transfer model and solid fuel chunk jamming model are developed and introduced to SAS4A, and validated using CABRI-2 E13 experimental data. Secondly, an investigation of the mechanism of energetics in the initiating phase of an unprotected loss-of-flow accident has identified major control parameters which are intimately related to core design parameters and material motion phenomena. (author)

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Residents within 30 km of the Sendai nuclear plant in Kagoshima prefecture, the scope of the evacuation zone, oppose to restart of the plant being shut down after Fukushima Daiichi nuclear accident. Findings based on the region's representative sample survey by interviewers

International Nuclear Information System (INIS)

Hirose, Hirotada

2015-01-01

Due to the aftermath of the accident of Tokyo Electric Power Company Right Single Quotation Marks Fukushima Daiichi Nuclear Power Station caused by the Great East Japan Earthquake, a lot of residents were still forced to be under evacuation. However, the Japanese government places nuclear power as a base load power, and is trying to successively restart nuclear power plants under full-stop state. Under such circumstance, a questionnaire survey about the issue over the nuclear power plant was conducted for residents near the Sendai Nuclear Power Station, which is expected to restart. The areas of the survey are within 30 km radius from the plant, where evacuation is required in case of an accident, and these areas are divided to two categories; Precautionary Action Zone (PAZ), which is within 5 km from a nuclear plant, and Urgent Protective Action Planning Zone (UPZ), which is 5-30 km from a nuclear plant. This paper introduces in detail the survey methods and findings. (A.O.)

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Superphenix set to rise again

International Nuclear Information System (INIS)

Dorozynski, A.

1993-01-01

Superphenix, France's seemingly jinxed fast breeder reactor, which has not produced a single kilowatt of energy in more than 3 years, looks set to rise up next year like the mythical bird it is named after. The $5 billion reactor, the largest fast breeder in the world, has just been given the seal of approval by a public commission ordered by the government to look at the pros and cons of restarting. It still has hoops to jump through: a safety check and approval from the ministries of industries and environment. But the consortium of French, Italian, and German power utilities that run the plant are confident they can get it running by next summer. The Superphenix that rises out of the ashes will, however, be a different species of bird from the one planned 20 years ago. The consortium plans to turn the reactor into a debreeder, one that will incinerate more plutonium than it produces and so eat into Europe's plutonium stockpile. Calculations by Superphenix staff and the Atomic Energy Commission indicate that a plutonivorous fast breeder could incinerate 15 to 25 kilograms of plutonium while producing 1 billion kilowatt-hours of electricity-scarcely enough to make a dent in the tonnes of plutonium produced by Electricite de France's reactors each year. The Superphenix consortium is anxious to get the reactor back on line. The annual cost of upkeep and repair of the idle plant and salaries for its 700 staff may reach $140 million this year, 20% more than if the plant was running normally. If restarted, the existing core and a second one ready on the shelf will generate electricity worth $1.3 billion

Phenix: a story of core and energy

International Nuclear Information System (INIS)

Sauvage, J.F.

2005-01-01

Phenix is the name of a legendary bird which could have several successive lives thanks to a rebirth from its ashes. It is by analogy the name given to an original sodium-cooled fast breeder reactor capable to generate new quantities of energy using the by products of its core burnup. This book tells the story of this reactor: construction (1968-1974), first years of operation (1974-1980), success era (1980-1986), first problems (1986-1992), safety re-evaluation (1992-1998), renovation of the core (1998-2003), re-start up of operation (2003-2009). A description of the power plant is given in appendix: core, reactor vessel, circuits, handling, instrumentation and control, safety, buildings, operation. (J.S.)

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

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

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

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

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

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

1996-01-01

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

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

Intercellular signaling via cyclic GMP diffusion through gap junctions restarts meiosis in mouse ovarian follicles.

Science.gov (United States)

Shuhaibar, Leia C; Egbert, Jeremy R; Norris, Rachael P; Lampe, Paul D; Nikolaev, Viacheslav O; Thunemann, Martin; Wen, Lai; Feil, Robert; Jaffe, Laurinda A

2015-04-28

Meiosis in mammalian oocytes is paused until luteinizing hormone (LH) activates receptors in the mural granulosa cells of the ovarian follicle. Prior work has established the central role of cyclic GMP (cGMP) from the granulosa cells in maintaining meiotic arrest, but it is not clear how binding of LH to receptors that are located up to 10 cell layers away from the oocyte lowers oocyte cGMP and restarts meiosis. Here, by visualizing intercellular trafficking of cGMP in real-time in live follicles from mice expressing a FRET sensor, we show that diffusion of cGMP through gap junctions is responsible not only for maintaining meiotic arrest, but also for rapid transmission of the signal that reinitiates meiosis from the follicle surface to the oocyte. Before LH exposure, the cGMP concentration throughout the follicle is at a uniformly high level of ∼2-4 μM. Then, within 1 min of LH application, cGMP begins to decrease in the peripheral granulosa cells. As a consequence, cGMP from the oocyte diffuses into the sink provided by the large granulosa cell volume, such that by 20 min the cGMP concentration in the follicle is uniformly low, ∼100 nM. The decrease in cGMP in the oocyte relieves the inhibition of the meiotic cell cycle. This direct demonstration that a physiological signal initiated by a stimulus in one region of an intact tissue can travel across many layers of cells via cyclic nucleotide diffusion through gap junctions could provide a general mechanism for diverse cellular processes.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

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

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

International Nuclear Information System (INIS)

Shin, Hyeong-Ki

1999-01-01

The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

RA Reactor

International Nuclear Information System (INIS)

1978-02-01

In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

Economic simplified boiling water reactor (ESBWR) response to an extended station blackout/ loss of all AC power

International Nuclear Information System (INIS)

Barrett, A.J.; Marquino, W.

2013-01-01

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackout for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event

H Reactor

Data.gov (United States)

Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

L-Lake zooplankton: L-Lake/Steel Creek Biological Monitoring Program, November 1985--December 1991

Energy Technology Data Exchange (ETDEWEB)

Bowers, J.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Bowen, M. [Normandeau Associates, Inc., New Ellenton, SC (United States)

1992-03-01

The L- Lake Biological Monitoring Program was designed to meet environmental regulatory requirements associated with the restart of L-Reactor and address portions of Section 316(a) of the Clean Water Act, which requires an applicant for a discharge permit to provide scientific evidence that the discharge causes no significant impact on the indigenous ecosystem. The Department of Energy (DOE) must demonstrate that the discharge of L-Reactor affluent into L Lake will not inhibit the eventual establishment of a ``Balanced Biological Community`` (BBC) in at least 50% of the lake. This report details results of monitoring zooplankton populations in L-Lake.

L-Lake zooplankton: L-Lake/Steel Creek Biological Monitoring Program, November 1985--December 1991

International Nuclear Information System (INIS)

Bowers, J.A.; Bowen, M.

1992-03-01

The L- Lake Biological Monitoring Program was designed to meet environmental regulatory requirements associated with the restart of L-Reactor and address portions of Section 316(a) of the Clean Water Act, which requires an applicant for a discharge permit to provide scientific evidence that the discharge causes no significant impact on the indigenous ecosystem. The Department of Energy (DOE) must demonstrate that the discharge of L-Reactor affluent into L Lake will not inhibit the eventual establishment of a ''Balanced Biological Community'' (BBC) in at least 50% of the lake. This report details results of monitoring zooplankton populations in L-Lake

550 tests to support Tihange 2 and Doel 3 integrity assessment. Extensive research based on SCK-CEN's long-term expertise

International Nuclear Information System (INIS)

2014-01-01

The article discusses the analyses that were performed in order to assess of the impact of the hydrogen flakes on the mechanical properties of the vessels of the Doel 3 and Tihange 2 nuclear power plants . Results from 550 tests shows no significant influence on the mechanical properties. Both reactors were given the green light by the regulator AFCN/FANC to restart operation. On specific request of the Belgian safety authorities, an interim research programme was set up in parallel to study the effects of irradiation on the material. The BR2 reactor is engaged for this research.

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

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

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

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

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Multiregion reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

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

Wolsong Unit 1 restart chemistry procedures during retubing outage

International Nuclear Information System (INIS)

Yun, Hyunran; Lee, Sarang; Moon, Yunyong; Kim, Seoyul

2015-01-01

Lay-up is aimed at protecting systems from degradation during outage, mainly by minimizing corrosion and particularly, when the outage is longer than 16 weeks. Due to the intrinsic design of CANDU reactors, their horizontal fuel channels should be replaced for another service life time. This poster presents the lay-up guidelines and methods recommended for re-tubing outage based on the first re-tubing operation made in Korea (at the Wolsung Unit 1). It is shown that dry lay-up with specific gas blanket was the sole choice for the primary heat transfer system, the moderator system and the steam cycle system while wet lay-up under circulation was recommended for the end shield cooling system and the liquid zone control system. The water filled part of steam generators, of the liquid zone control system and of the end shield cooling system was maintained normal

The uranium market: 1986-2000

International Nuclear Information System (INIS)

Lewiner, C.; Walton, D.; Sinclair-Smith, D.

1987-01-01

This paper summarizes the main conclusions of the 1986 supply and demand report of the Uranium Institute. The probable nuclear generating capacity is estimated for 1986-2000. Previous capacity forecasts (1981-1986) are compared with actual generating capacity. When looking at demand, a distinction has to be made between what reactor operators require to fuel reactors (reactor requirements) and what they intend to purchase (uranium procurements). This distinction is defined and discussed. The interaction between supply and demand is shown and factors affecting trade (eg government policies) are discussed. 1985 was the first year when uranium production was less than reactor requirements. This shortfall will continue for a number of years with the current excess in consumer inventories supplying the difference between reactor requirements and uranium procurements. Uranium demand should exceed production by 1988 but additional capacity should be available from planned and possible restart and possible new facilities. (U.K.)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it