Human-system interfaces for the ABWR control room

International Nuclear Information System (INIS)

Gutierrez, R.; O'Neil, T.J.

1993-01-01

The 1300-MW(electric) advanced boiling water reactor (ABWR) design has been developed by General Electric Co. (G.E.) and its technical associates, Hitachi Ltd. and Toshiba Corporation, under the sponsorship of the Tokyo Electric Power Company. Key features of the ABWR include simplification, improved safety and reliability, reduced occupational exposure and radwaste, improved maneuverability, and reduced construction, fuel, and operating costs. The ABWR incorporates the best proven features from BWR designs in Japan, Europe, and the United States, combined with the broad-scope application of leading-edge technology. The first application of the ABWR design will be at TEP-CO's Kashiwazaki-Kariwa nuclear power station units 6 and 7, which are scheduled to begin commercial operation in 1996 and 1997, respectively. The ABWR is also the lead plant in the U.S. standard plant design certification program. It is scheduled to receive final approval from the U.S. Nuclear Regulatory Commission in early 1994. The development of the human system interaces (HSI) for the ABWR is an extension of the operating experience of existing man-machine interfaces currently being used at GE-designed plants

Outline of the advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

Hucik, S.A.; Imaoka, T.; Minematsu, A.; Takashima, Y.

1986-01-01

The fundamental design of the Advanced Boiling Water Reactor (ABWR) was completed in December 1985. This design represents the next generation of Boiling Water Reactors (BWR) to be introduced into commercial operation in the 1990s. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technologies and many new improvements based on an extensive accumulation of world-wide experience through design, construction and operation of BWRs. The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. The ABWR objectives were specific improvements such as operating and safety margins, enhanced availability and capacity factor, and reduced occupational exposure while at the same time achieving significant cost reduction in both capital and operating costs. The ABWR is characterized by an improved NSSS including ten internal recirculation pumps, fine motion electric-hydraulic control rod drives, optimized safety and auxiliary systems, advanced control and instrumentation systems, improved turbine-generator with moisture/separator reheater with plant output increased to 1350 MWe, and an integrated reinforced concrete containment vessel and compact Reactor and Turbine Building design. The turbine system also included improvements in the Turbine-Generator, feedwater/heater system, and condensate treatment systems. The radwaste system was also optimized taking advantage of the plant design improvements and advances in radwaste technology. The ABWR is a truly optimal design which utilizes advanced technologies, capabilities, performance improvements, and yet provides an economic advantage. (author)

Study of advanced LWR cores for effective use of plutonium and MOX physics experiments

International Nuclear Information System (INIS)

Yamamoto, T.; Matsu-Ura, H.; Ueji, M.; Ota, H.; Kanagawa, T.; Sakurada, K.; Maruyama, H.

1999-01-01

Advanced technologies of full MOX cores have been studied to obtain higher Pu consumption based on the advanced light water reactors (APWRs and ABWRs). For this aim, basic core designs of high moderation lattice (H/HM ∼5) have been studied with reduced fuel diameters in fuel assemblies for APWRs and those of high moderation lattice (H/HM ∼6) with addition of extra water rods in fuel assemblies for ABWRs. The analysis of equilibrium cores shows that nuclear and thermal hydraulic parameters satisfy the design criteria and the Pu consumption rate increases about 20 %. An experimental program has been carried out to obtain the core parameters of high moderation MOX cores in the EOLE critical facility at the Cadarache Centre as a joint study of NUPEC, CEA and CEA's industrial partners. The experiments include a uranium homogeneous core, two MOX homogeneous cores of different moderation and a PWR assembly mock up core of MOX fuel with high moderation. The program was started from 1996 and will be completed in 2000. (author)

ABWR evolution program

International Nuclear Information System (INIS)

Omoto, A.; Tanabe, A.; Moriya, K.; Dillmann, C.W.

1993-01-01

The ABWR plant is becoming a commercial reality in Japan where the first two units are being built by the Tokyo Electric Power Company. Although these units are scheduled to come on line in 1996 and '97, the Tokyo Electric Power Company, jointly with NSSS vendors (General Electric, Toshiba and Hitachi) and Japanese BWR utilities, initiated a program for a new plant design. This program is aimed at the further evolution of the ABWR to take advantage of new technological developments and to meet possible social changes in the years to come. The expected time for the first-of-a-kind plant to come on line is in the 2010's when the first generation plants in Japan may approach the time for replacement. This paper presents the ouline of this program with focus on the utility requirements and candidate technologies. (orig.)

Development of advanced BWR fuel bundle with spectral shift rod (3) -transient analysis of ABWR core with SSR

International Nuclear Information System (INIS)

Ikegawa, T.; Chaki, M.; Ohga, Y.; Abe, M.

2010-01-01

The spectral shift rod (SSR) is a new type of water rod, utilized instead of the conventional water rod, in which a water level develops during core operation. The water level can be changed according to the fuel channel flow rate. In this study, ABWR plant performance with SSR fuel bundles under transient conditions has been evaluated using the TRACG code. The TRACG code, which can treat three-dimensional hydrodynamic calculations in a reactor pressure vessel, is well suited for evaluating the reactor transient performance with the SSR fuel bundles because it can calculate the water levels in the SSR at each channel grouping and therefore evaluate the core reactivity according to the water level changes in the SSR. 'Generator load rejection with total turbine bypass failure' and 'Recirculation flow control failure with increasing flow' were selected as cases which may increase the reactivity with the increasing water level in the SSR. It was found that the absolute value of the void reactivity coefficient in the SSR core was larger than that in the conventional water rod core because the core averaged void fraction in the SSR core, which has the vapor region above the water level in the SSR, was larger than that in the conventional water rod core. Therefore, AMCPR for the SSR core was a little larger than that for the conventional water rod core; however, the difference was smaller than 0.02 because the inlet of the SSR ascending path was designed to be small enough to prevent the rapid water level increase in the SSR. (authors)

Features of ABWR operator training with a full-scope simulator

International Nuclear Information System (INIS)

Kondou, Shin'ichi

1999-01-01

Many innovations have been incorporated into the Advanced BWR (ABWR) type control panels. In the BWR Operator Training Center (BTC), we started ABWR operator training using an ABWR full-scope simulator prior to the first ABWR plant's commercial operation. In consideration of the features of the ABWR type control panels, BTC has been conducting ABWR operator training focusing on the following 2 points; (1) Operator training reflecting the differences in the Human-Machine Interface (HMI). The new HMI devices which have the touch-operation function were introduced. These devices have higher operability, however, they require new operational skills. We planned the training program so that operators can fully acquire these skills. Also the compact main console and the new HMI devices made it relatively difficult for the operator crews to grasp visually what an operator was doing. We provide the training to have proper communication skills, and check trainees' operation using monitoring systems for simulator training. (2) Operator training responding to the expanded operation automation system. The scope of the automation system was expanded to reduce the operators' burden. We provide the training to improve the trainees' competence for 'operation and monitoring' suitable to both manual and automatic operational modes. (author)

The status of development activities of ABWR-2

International Nuclear Information System (INIS)

Anegawa, T.; Aoyama, M.; Fennern, L.E.; Koh, F.

2001-01-01

This paper reports the current status of development activities for ABWR-II which is the next generation design based on ABWR. Six Japanese BWR utilities and three BWR plant venders, General Electric Company, Hitachi Ltd. and Toshiba Corporation, have jointly been promoting the program. The program has been progressing for a decade. In Phase I (1991-92), future technologies were discussed and several plant concepts were studied. In Phase II (1993-95), in order to establish a reference reactor concept, key design features were selected. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements have been done in order to satisfy the design requirements well. The start of commercial operation of ABWR-II is estimated to be in the late 2010 when replacements of the first generation of operating nuclear power plants are expected to start. Larger power output and some passive safety features are adopted in order to make ABWR-II competitive in power generation cost and attain public confidence on safety. The advantages of a large fuel bundle design are also discussed here. (author)

Alarm system for ABWR main control panels

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Yuji; Saito, Koji [Toshiba Corp., Yokohoma (Japan)

1997-09-01

TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ``Alarm System for ABWR Main Control Panels``. (author). 5 refs, 8 figs, 1 tab.

Alarm system for ABWR main control panels

International Nuclear Information System (INIS)

Kobayashi, Yuji; Saito, Koji

1997-01-01

TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ''Alarm System for ABWR Main Control Panels''. (author). 5 refs, 8 figs, 1 tab

Application of ABWR construction database to nuclear power plant project

International Nuclear Information System (INIS)

Takashima, Atsushi; Katsube, Yasuhiko

1999-01-01

Tokyo Electric Power Company (TEPCO) completed the construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and No. 7 (K-6/7) as the first advanced boiling water reactors (ABWR) in the world successfully. K-6 and K-7 started their commercial operations in November, 1996 and in July, 1997 respectively. We consider ABWR as a standard BWR in the world as well as in Japan because ABWR is highly reputed. However, because the interval of our nuclear power plant construction is going to be longer, our engineering level on plant construction will be declining. Hence it is necessary for us to maintain our engineering level. In addition to this circumstance, we are planning to wide application of separated purchase orders for further cost reduction. Also there is an expectation for our contribution to ABWR plant constructions overseas. As facing these circumstances, we have developed a construction database based on our experience for ABWR construction. As the first step of developing the database for these use, we analyzed our own activities in the previous ABWR construction. Through this analysis, we could define activity units of which the project consists. As the second step, we clarified the data which are treated in each activity unit and the interface among them. By taking these steps, we could develop our database efficiently. (author)

Further improvement of human-machine interface for ABWR main control room

International Nuclear Information System (INIS)

Makino, S.

2001-01-01

Tokyo Electric Power Company (TEPCO) has developed main control room panels based on progress in C and I technology. ABWR type main control room panels (ABWR MCR PNLs) are categorized as third generation type domestic BWR MCR, that is, they are were developed step by step based on operating experience with the first and the second generation BWR. ABWR type main control room panels were applied to Kashiwazaki-Kariwa Nuclear Power Station Units Number 6 and 7 (K-6/7) for the first time. K-6/7 are the first advanced BWR (ABWR), which started commercial operation in November 1996 and July 1997, respectively. The concept of ABWR MCR design was verified through wooden mock-up panels, start-up tests and commercial operation. Though the K-6/7 design has borne fruit, we are planning to refine and standardize the design based on the following concepts: to maintain the plant operation and monitoring style of ABWR MCR PNLs; to introduce brand-new HMI technology and devices; to incorporate operators' advice in the design. This paper outlines the features and improvements of the K6/7 MCR PNLs design. (author)

Advanced boiling water reactor (ABWR). Design, construction, operation and maintenance experience

International Nuclear Information System (INIS)

Idesawa, M.

1998-01-01

The ABWR has experienced all phases of design, construction, operation and maintenance at Kashiwazaki-Kariwa Nuclear Power Station Units No.6 and 7 and confirmed that originally intended development targets have been achieved with highly satisfactory results. This is the fruit of a project that collected wisdom from various sources under a international cooperative organization, with Tokyo Electric Power Company taking the leading role from the onset. These two units have not only demonstrated that ABWRs have superior performance as the first standard units of advanced light water reactor but also aroused a hope for the big potential advantages that ABWRs can provide us. The ABWR has already been awarded a U.S. standard license for having proved that it can comply with the requirements of international regulatory systems with an ample margin. There are also many construction programs with ABWRs progressing both domestically and abroad, suggesting that it has won recognition as an international standard plant. We will do our utmost to perfect the operation and maintenance records of Kashiwazaki-Kariwa Units No.6 and 7, which is the top runner among ABWRs, and to make known the superiority of this reactor to the world. (J.P.N.)

Cost reduction and safety design features of ABWR-II. Annex 5

International Nuclear Information System (INIS)

Koh, F.; Moriya, K.; Anegawa, T.

2002-01-01

The ABWR-II, which is aimed to be the next generation reactor following the latest BWR: Advanced Boiling Reactor (ABWR), is now under development jointly by the Japanese BWR utilities, General Electric Company, Hitachi Limited, and Toshiba Corporation. The key objectives of ABWR-II development include improvement in economics and further sophistication in safety for commercialization in the late 2010's and after. This paper summarizes the current status of ABWR-II development focusing on economics and safety. Plant power rating, fuel size, CRD rationalization and outage period are discussed from a cost reduction perspective. In terms of safety, the features such as diversification in emergency power sources and passive system application against severe accidents are being introduced. (author)

Development, operating experience and future plan of ABWR in Japan

International Nuclear Information System (INIS)

Ujihara, N.

2001-01-01

From 1974 to 1978 nuclear power plants' capacity factor decreased to a low level at the Tokyo Electric Power Company (TEPCO) for stress corrosion cracking of stainless steal pipe, thermal fatigue crack of feed water spurger and vibration trouble of LPRM, etc. So, for improving that situation, we gathered the most useful BWR technology in the world at that time and we started the ABWR project to develop an ideal BWR. The goals of this development were: (1) Enhanced safety and reliability; (2) Reduced occupational radiation exposure and radioactive waste; (3) Enhanced operability and maneuverability; and (4) Improved economy. Significant improvements were induced by the adoption of the reactor internal pump, fine motion control rod drive mechanism, integral type reinforced concrete containment vessel and digital control system. After about 10 years R and D, we succeeded in the development of the compact and economic plant, ABWR (1356MW(e)) whose building volume had been reduced to 76%, compared with the previous BWR (1100MW(e)). The construction of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6, which is the first ABWR plant in the world, was started September 1991, and that of unit No. 7 was started in February 1992. We used effective construction methods, such as large block construction method and all weather construction method, to shorten the construction duration from first concrete to commercial operation start to 51.5 months. The commercial operation start of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6 was November 1996 and that of unit No. 7 was July 1997. After that, unit No. 6 had three times the annual outage and unit No. 7 had two times the annual outage. Both plants have good operating experiences. From commercial operation start to end of August 2000, the capacity factor of unit No. 6 is 86.4% and that of unit No. 7 is 86.9%. Now, in Taiwan, ABWR plants are under construction at Lungmen unit No. 1 and 2. In Japan, ABWR plants are under

MELCOR 1.8.2 calculations of selected sequences for the ABWR

International Nuclear Information System (INIS)

Kmetyk, L.N.

1994-07-01

This report summarizes the results from MELCOR calculations of severe accident sequences in the ABWR and presents comparisons with MAAP calculations for the same sequences. MELCOR was run for two low-pressure and three high-pressure sequences to identify the materials which enter containment and are available for release to the environment (source terms), to study the potential effects of core-concrete interaction, and to obtain event timings during each sequence; the source terms include fission products and other materials such as those generated by core-concrete interactions. Sensitivity studies were done on the impact of assuming limestone rather than basaltic concrete and on the effect of quenching core debris in the cavity compared to having hot, unquenched debris present

Man-machine interface systems and operator training program for ABWR in Japan

International Nuclear Information System (INIS)

Kunito, Susumu

2004-01-01

The Tokyo Electric Power Company (TEPCO) has developed a new Main Control Room design for the Advanced Boiling Water Reactor (ABWR) to improve man-machine interface. New configuration of panels and enhanced automation are some of the features of the ABWR type Main Control Room design. Various technologies such as Cathode Ray Tubes (CRTs) and Flat Displays (FDs) with touch-sensitive operations are contributed to the development of the ABWR type control room design. This design will be first applied to Kashiwazaki-Kariwa Nuclear Power Station unit 6 (K-6). To train the operators sufficiently, TEPCO reviewed the operator training program. Compared with the conventional training, new training menu will be added and the training of ABWR operators will be started 6 months earlier. An ABWR simulator is under construction and training using this simulator is scheduled to be started in August 1994, which is 18 months before fuel loading of K-6. We are reviewing malfunction modes on the simulator. (author)

U.S. licensing process and ABWR certification

International Nuclear Information System (INIS)

Quirk, J.F.; Williams, W.A.

1996-01-01

Part 50 of Title 10 of the Code of Federal Regulation (CFR) establishes a two-step licensing process by which the U.S. Nuclear Regulatory Committee (NRC) authorizes nuclear reactor plant construction through issuance of a construction permit and authorizes operation by issuance of an operating license. At each stage, the NRC Staff conducts technical reviews and there is potential for public hearings. In 1989, the NRC issued a new, simplified licensing process: Part 52. The purpose of the Part 52 licensing process is to provide a regulatory framework that brings about earlier resolution of licensing issues. Because issues are not resolved early in the Part 50 licensing process, approval of an operating license is not assured until after a significant investment has been made in the plant. Part 52 increases the stability and certainty of the licensing process by providing for the early resolution of safety and environmental issues. The Part 52 licensing process features (1) early site permits, (2) design certification, and (3) combined construction permit and operating licenses. As part of the U.S. Advanced Light Water Reactor (ALWR) Program to revitalize the nuclear option through the integration of government/utility/industry efforts, GE undertook the role of applying for certification for its latest product line, the Advanced Boiling Water Reactor (ABWR), under the U.S. ABWR certification program. The ABWR design is an essentially complete plant. Initial application for design certification was in 1987 under Part 50. GE reapplied in late 1991 under the newly promulgated Part 52. Following seven years of intensive interactions with the NRC and ACRS, GE was awarded the first Final Design Approval (FDA) under Part 52. The Commission initiated rulemaking by publishing the proposed ABWR Certification Rule in the Federal Register in early 1995. Certification is anticipated mid-1996. (J.P.N.)

What kind of nuclear energy should be offered? Ge's ABWR nuclear plant

International Nuclear Information System (INIS)

Hucik, S.; Redding, J.

2000-01-01

GE is proud of the ABWR design and our accomplishments in providing this safe, cost effective means of generating electricity to our customers around the world. The ABWR defines the new generation of advanced nuclear plant designs that have achieved higher levels of safety, compete economically with other forms of power generation, and can contribute importantly to the goals of sustainable development. The GE-led team has amassed significant experience in licensing, designing, and constructing the ABWR and has established a strong track record of success in doing so. We believe the quality of the ABWR design and our experience will have a strong appeal to utility business managers and the financial community. Finally we believe in nuclear energy and it's ability to improve the quality of people's lives throughout the world. Nuclear energy can and should play an important role in meeting the worlds' growing needs for electricity while at the same time preserving our environment for future generations. (authors)

Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

International Nuclear Information System (INIS)

Kamei, Kazuhiro; Tanoue, Tetsuharu; Kataoka, Kazuyoshi; Jimbo, Masakazu

2011-01-01

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

Simulation technology to develop the automatic control rod operation system in ABWR

Energy Technology Data Exchange (ETDEWEB)

Yoshihiko, Ishii; Atsushi, Fushimi; Naoyuki, Ishida [Hitachi Ltd. Power and Industrial Systems R and D Laboratory, Ibaraki (Japan); Hitoshi, Ochi; Masakazu, Yuzuki [Hitachi Ltd. Power Systems, Hitachi Works, Ibaraki (Japan); Hideki, Hanami [Hitachi Ltd. Information and Control System Div., Ibaraki (Japan)

2007-07-01

To design and verify the automatic power regulator system (APR) for an advanced boiling water reactor (ABWR), Hitachi has developed a plant simulation code system. The 3-dimensional kinetic analysis code STAND, based on a polynomial nodal expansion method, was effective in developing an APR automatic control rod operation algorithm in the critical mode. The point kinetic plant simulator AUTSIMU is an off-line and real time online plant simulation code that can simulate ABWR start-up and shutdown processes from a sub-critical condition with a water temperature of 20 degrees C to the rated operation power level at about 7 MPa. APR operation algorithms were designed using the off-line mode. In this mode, less than one minute in real time was needed to simulate a ten-hour transient with an ordinary personal computer. In the online mode, APR hardware was efficiently verified in a factory test before shipping. The codes were verified with ABWR start-up experimental data. It was confirmed that the simulators were effective in simulating ABWR plant start-up behavior. (authors)

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-04-30

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

ABWR1. A Generation III.7 reactor after the Fukushima Daiichi accident

International Nuclear Information System (INIS)

Sato, Takashi; Matsumoto, Keiji; Kurosaki, Toshikazu; Taguchi, Keisuke

2015-01-01

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. The NSSS is exactly the same as that of the current ABWR. As for safety design it has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7 day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the CV building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most

A study on transient heat transfer of the EU-ABWR external core catcher using the phase-change effective convectivity model

International Nuclear Information System (INIS)

Tran Chi Thanh; Nguyen Viet Hung; Tahara, Mika; Kojima, Yoshihiro; Hamazaki, Ryoichi; Kudinov, Pavel

2015-01-01

In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation III and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and

High moderation MOX cores for effective use of plutonium in LWRs

International Nuclear Information System (INIS)

Hamamoto, Kazuko; Kanagawa, Takashi; Hiraiwa, Koji; Sakurada, Koichi; Moriwaki, Masanao; Aoyama, Motoo; Yamamoto, Toru; Ueji, Masao

2001-01-01

Conceptual design studies have been performed for high moderation full MOX cores aiming at increasing fissile Pu consumption rate (ratio of the consumed to the loaded fissile Pu) and reducing residual Pu in discharged MOX fuel. The BWR cores studied have hydrogen to heavy metal ratio(H/HM) of 5.9 with increasing water rods and 7.0 with reducing a fuel rod diameter based on a reference 9x9 fuel (H/HM=4.9) of ABWR. The PWR cores studied have H/HM of 5.0 and 6.0 with reducing a fuel rod diameter based on a reference 17x17 fuel (H/HM=4.0) of APWR. Equilibrium core design and plant safety analyses showed that those high moderation cores have compatibility with ABWR and APWR. The fissile Pu consumption rate is 22% larger than the full MOX cores with reference fuel of ABWR and 50% for APWR. The core performance and compatibility has been also evaluated in the condition of multi-recycle of Pu in these high moderation cores. Study has been conducted to evaluate the effect of introducing these high moderation cores in the fuel cycle of Japan. It shows that the high moderation cores produce 26% more cumulative electricity and reduce 22% stock of the fissile Pu by 2050 than the reference cores. (author)

GE-Hitachi ABWR Design for 60 Years of Operation

International Nuclear Information System (INIS)

Hoang Xuan Hoa

2011-01-01

The GE Hitachi ABWR design benefits from the experiences of the prior generation BWRs in operation since the 1970s, and the knowledge learned from years of technical research on the aging effects on reactor systems, structures and components. The impact of exposure to neutrons over time to the reactor pressure vessel and internal components was evaluated to confirm their structural integrity. Other time-based plant analyses, such as fatigue vibration and corrosion effect, have also been performed on the ABWR to support a 60-year plant life assumption. Finally, operating experience and aging lessons learned have been used to verify the maintenance plans for plant aging management are acceptable to support operation up to 60 years. (author)

Development of ABWR-2 and its safety design

International Nuclear Information System (INIS)

Takafumi, Anegawa; Kenji, Tateiwa

2002-01-01

This paper reports the current status of development project on ABWR-II, a next generation reactor design based on ABWR, and its safety design. This project was initiated over a decade ago and has completed three phases to date. In Phase I (1991-92), basic design requirements were discussed and several plant concepts were studied. In Phase II (1993-95), key design features were selected in order to establish a reference reactor concept. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements were made to fulfill the design requirements. By adopting large electric output (1 700 MW), large fuel bundle, modified ECCS, and passive heat removal systems, among other design features, we achieved a design concept capable of increasing both economic competitiveness and safety performance. Main focus of this paper will be on the safety design, safety performance, and further research needs related to safety. (authors)

Core design study on reduced-moderation water reactors

International Nuclear Information System (INIS)

Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

2002-01-01

The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

ABWR certification work brings US licensing stability nearer

International Nuclear Information System (INIS)

Wilkins, D.R.; Quirk, J.F.

1991-01-01

The Advanced Boiling Water Reactor (ABWR) is now approaching Final Design Approval by the US Nuclear Regulatory Commission (NRC) and will then proceed on to the certification phase of the NRC's new standard plant licensing process. Successful completion of this will usher in a new era of standardization and reactor licensing stability in the US. (author)

Design features which mitigate severe accident challenges in the GE ABWR and SBWR

International Nuclear Information System (INIS)

Buchholz, Carol E.

2004-01-01

A reduction of the requirements for the emergency planning zone (EPZ) is a goal of advanced light water reactors. The technical basis for reducing the EPZ requirements is based on a very low frequency of a severe accident and high confidence that the offsite dose would be low even if a severe accident was to occur. Design features have been included in both the ABWR and SBWR to ensure that both of these goals are achieved. Probabilistic Risk Assessments (PRAs) have been performed for both plants. The PRAs indicate a core damage frequency on the order of IE-7 for both plants. The PRAs also show that the containments will not fail even if a severe accident should occur. The potential offsite is extremely low. (author)

A passive decay-heat removal system for an ABWR based on air cooling

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

2017-01-15

Highlights: • A passive decay heat removal system for an ABWR is discussed using combined system of the reactor and an air cooler. • Effect of number of pass of the finned heat transfer tubes on heat removal is investigated. • The decay heat can be removed by air coolers with natural convection. • Two types of air cooler are evaluated, i.e., steam condensing and water cooling types. • Measures how to improve the heat removal rate and to make compact air cooler are discussed. - Abstract: This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as

Overview of internal fire hazards aspects of ABWR design for United Kingdom

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Kazuhiro; Kawai, Hiroki [Hitachi-GE Nuclear Energy, Ltd., Ibaraki (Japan)

2015-12-15

The ABWR (Advanced Boiling Water Reactor) is a generation III+ reactor, the most modern operational generation of nuclear power plants. The UK ABWR design is proposed for development and construction in the United Kingdom (UK), and under review by the Office for Nuclear Regulation (ONR) through Generic Design Assessment (GDA). The UK ABWR design has mainly two types of the safety system: ''preventing'' and ''mitigating'' a fault and their consequences. The prevention of internal hazards starts with design processes and procedures. These processes lead to limiting the sources of potential hazards. The mitigative safety systems are required to ensure the fundamental safety functions (FSFs): control of reactivity, Fuel cooling, long term heat removal, confinement/containment of radioactive materials, and others. Implementation of the safety philosophy is based upon redundant and diverse safety systems that deliver the FSFs. Three mechanical divisions are provided, each of which contains redundant systems, structures, and components (SSCs) capable of carrying out all the FSFs. The safety divisions are separated by robust barriers which act to contain a hazard in an affected division and prevent the spread of the hazard to a different division. The deterministic assessments and the hazard schedule argue that the rooms containing SSCs providing the FSFs are located in different fire safety divisions. The approach to maintaining the FSFs during and after internal fires is to ensure fires do not spread beyond that division to affect redundant equipment in other divisions. During the GDA process, it is demonstrated that generally barrier compartmentation (the divisional barrier walls, ceilings and floors) is sufficient to contain the postulated fires. The UK ABWR design has sufficient capability of withstanding the postulated internal fire hazard to achieve the FSFs. Further development is being undertaken with feedback in the GDA

Study of plutonium multi-recycle in high moderation LWR cores

International Nuclear Information System (INIS)

Iwata, Yutaka; Yamamoto, Toru; Ueji, Masao; Hibi, Koki; Aoyama, Motoo; Sakurada, Koichi

2000-01-01

Nuclear Power Engineering Corporation (NUPEC) has been studying advanced cores that are dedicated to enhance the plutonium consumption per recycling for effective use of plutonium. In this study, a fissile plutonium consumption rate is adopted as an index of the effective use of plutonium, which is defined as a ratio of consumption to loading of fissile plutonium in a core. High moderation core concepts have been studied in order to increase this index based on full MOX cores in the latest designs of LWRs in Japan that are the Advanced Boiling Water Reactor (ABWR) and the Advanced Pressurized Water Reactor (APWR). As a part of this study, core performance in the case of plutonium multi-recycling has been surveyed with these higher moderation cores aiming further effective use of plutonium. The design and analyses for equilibrium cores show that nuclear and thermal hydraulics parameters satisfy design criteria, and a fissile plutonium consumption rate increases up to 20% for ABWRs and 30% for APWRs even in plutonium multi-recycling condition. It was confirmed that the high moderation cores are feasible from a viewpoint of nuclear and thermal hydraulics, safety and plutonium consumption in the condition of plutonium multi-recycling. (author)

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

Science.gov (United States)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

Transient analysis of ABWR reactor using a best estimate code

International Nuclear Information System (INIS)

Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.

2004-01-01

Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)

Upgraded prototype-reactor internal pump for ABWR

International Nuclear Information System (INIS)

Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

1988-01-01

In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

Experimental study on RCCV of ABWR plant-Part 1

International Nuclear Information System (INIS)

Saita, H.; Kikuchi, R.; Muramatsu, Y.; Hiramoto, M.; Oyamada, O.; Furukawa, H.; Sasagawa, K.; Ohmori, N.; Suzuki, S.; Suoita, M.; Kobayashi, I.; Yamaguchi, I.

1989-01-01

Steel containment vessels have generally been used for BWR (Boiling Water Reactor)-type nuclear power plants in Japan. However, in accordance with the development of the advanced-type BWR (ABWR), the introduction of a reinforced concrete containment vessel (RCCV) has been proposed because it enables greater freedom in flexibility of configuration. The study presented here accumulates basic data on RCCV's and verifies the integrity and ultimate strength of a trial-designed RCCV and the validity of its design method

An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

2012-09-01

To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

International Nuclear Information System (INIS)

Yang, C.-Y.; Liang, Thomas K.S.; Pei, B.S.; Shih, C.K.; Chiang, S.C.; Wang, L.C.

2009-01-01

For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its' own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.

Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

Seismic margin analysis for Kashiwazaki Kariwa ABWR plant considering the Niigataken Chuetsu-oki earthquake

International Nuclear Information System (INIS)

Matsuo, Toshihiro; Nagasawa, Kazuyuki; Kawamura, Shinichi; Ueki, Takashi; Higuchi, Tomokazu; Sakaki, Isao

2009-01-01

Seismic Margin Analysis (SMA) study was conducted for Kashiwazaki Kariwa (KK) ABWR representative plant (unit 6). Considering that the installation behaved in a safe manner during and after the Niigataken Chuetsu-oki (NCO) Earthquake which significantly exceeded the level of the seismic input taken into account in the design of the plant, the study to find out how much margin the ABWR plant had toward the same seismic motion was conducted. In this study fragility analyses were conducted for SSCs that were included in the accident sequences and that were considered to have relatively small margin taking EPRI margin analysis method into consideration. In order to calculate plant level seismic margin Min-Max method was adopted. As the result of this study, the plant level High Confidence Low Probability of Failure (HCLPF) acceleration for unit 6 was calculated more than tripled NCO earthquake motion. (author)

Probabilistic based design rules for intersystem LOCAS in ABWR piping

International Nuclear Information System (INIS)

Ware, A.G.; Wesley, D.A.

1993-01-01

A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures

Software failure events derivation and analysis by frame-based technique

International Nuclear Information System (INIS)

Huang, H.-W.; Shih, C.; Yih, Swu; Chen, M.-H.

2007-01-01

A frame-based technique, including physical frame, logical frame, and cognitive frame, was adopted to perform digital I and C failure events derivation and analysis for generic ABWR. The physical frame was structured with a modified PCTran-ABWR plant simulation code, which was extended and enhanced on the feedwater system, recirculation system, and steam line system. The logical model is structured with MATLAB, which was incorporated into PCTran-ABWR to improve the pressure control system, feedwater control system, recirculation control system, and automated power regulation control system. As a result, the software failure of these digital control systems can be properly simulated and analyzed. The cognitive frame was simulated by the operator awareness status in the scenarios. Moreover, via an internal characteristics tuning technique, the modified PCTran-ABWR can precisely reflect the characteristics of the power-core flow. Hence, in addition to the transient plots, the analysis results can then be demonstrated on the power-core flow map. A number of postulated I and C system software failure events were derived to achieve the dynamic analyses. The basis for event derivation includes the published classification for software anomalies, the digital I and C design data for ABWR, chapter 15 accident analysis of generic SAR, and the reported NPP I and C software failure events. The case study of this research includes: (1) the software CMF analysis for the major digital control systems; and (2) postulated ABWR digital I and C software failure events derivation from the actual happening of non-ABWR digital I and C software failure events, which were reported to LER of USNRC or IRS of IAEA. These events were analyzed by PCTran-ABWR. Conflicts among plant status, computer status, and human cognitive status are successfully identified. The operator might not easily recognize the abnormal condition, because the computer status seems to progress normally. However, a well

Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

Energy Technology Data Exchange (ETDEWEB)

Tzang, Y.-C., E-mail: yctzang@aec.gov.t [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China); Chiang, R.-F.; Ferng, Y.-M.; Pei, B.-S. [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China)

2009-12-15

In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the DELTACPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

ABWR (K-6/7) construction experience (computer-based safety system)

International Nuclear Information System (INIS)

Yokomura, T.

1998-01-01

TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)

Development of automated ultrasonic device for in-service inspection of ABWR pressure vessel bottom head

International Nuclear Information System (INIS)

Kojima, Y.; Matsuyama, A.

1995-01-01

An automated device and its controller have been developed for the bottom head weld examination of pressure vessel of Advanced Boiling Water Reactor (ABWR). The internal pump casings and the housings of control rod prevent a conventional ultrasonic device from scanning the required inspection zone. With this reason, it is required to develop a new device to examine the bottom head area of ABWR. The developed device is characterized by the following features. (1) Composed of a mother vehicle and a compact inspection vehicle. They are connected only by an electric wire without using the conventional arm mechanism. (2) The mother vehicle travels on a track and lift up the inspection vehicle to the vessel. (3) The mother vehicle can automatically attach the inspection vehicle to the bottom head, and detach the inspection vehicle from it. (4) Collision avoidance control function with a touch sensor is installed at the front of the inspection vehicle. The device was successfully demonstrated using a mock-up of reactor pressure vessel

HTR core physics analysis at NRG

International Nuclear Information System (INIS)

Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

2002-01-01

Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

The development and assessment of TRACE/PARCS model for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Lin, Hao-Tzu, E-mail: htlin@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County, 32546, Taiwan, ROC (China); Wang, Jong-Rong, E-mail: jrwang@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County, 32546, Taiwan, ROC (China); Chen, Hsiung-Chih, E-mail: km8385@gmail.com [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang Fu Road, HsinChu 30013, Taiwan, ROC (China); Shih, Chunkuan, E-mail: ckshih@ess.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang Fu Road, HsinChu 30013, Taiwan, ROC (China)

2014-07-01

Highlights: • The development of a TRACE/PARCS model of Lungmen (ABWR). • Thermal-hydraulics model of TRACE is coupling with 3-D neutronics model of PARCS. • The 3-D geometry vessel component of TRACE is used in this study. • The parameters responses of Lungmen TRACE/PARCS model are consistent with FSAR data. - Abstract: This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass and pressure control system, and recirculation flow control system), etc. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The transient data of Final Safety Analysis Report (FSAR) chapter 15th were used to compare with the results of Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the FSAR data. It indicated that there was a respectable accuracy in the Lungmen NPP TRACE/PARCS model and it also depicted that the Lungmen NPP TRACE/PARCS model was satisfying for the purpose of Lungmen NPP safety analyses.

Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

International Nuclear Information System (INIS)

Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

2012-01-01

Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

Development of ABWR inertia-increased reactor internal pump and thicker sleeve nozzle

International Nuclear Information System (INIS)

Takahashi, Shirou; Shiina, Kouji; Matsumura, Seiichi

2002-01-01

The conventional reactor internal pumps (RIPs) in the ABWR have an inertia moment coming from the shafts and Motor-Generator sets, enabling the RIPs to continue running for a few seconds, when a trip of all RIPs event occurs. It is possible to simplify the RIPs' power supply system without affecting the core flow supply when the above event occurs by eliminating M-G sets, if the rotating inertia is increased. This inertia increase due to an additional flywheel, which leads to gains in weight and length, requires the larger diameter nozzle with the thicker sleeve. However, too large a nozzle diameter may change the hydraulic performance. In authors' previous study, the optimum nozzle diameter (492 mm) was selected through 1/5-scale test. In this study, the 492 mm nozzle and the inertia-increased RIP were verified through the full-scale tests. The rotating inertia time constant on coastdown characteristics (behavior of the RIP speed in the event of power loss) for the inertia-increased RIP doubled compared with the current RIP. The casing and the shaft vibration were also confirmed to satisfy the design criteria. Moreover, hydraulic performance and heat increase in the motor casing due to the flywheel were evaluated. The inertia increased RIP with the 492 mm nozzle maintained good performance. (author)

Commercial operation and outage experience of ABWR at Kashiwazaki-Kariwa units Nos. 6 and 7

International Nuclear Information System (INIS)

Anahara, N.; Yamada, M.; Kataoka, H.

2000-01-01

Kashiwazaki-Kariwa Nuclear Power Station Units Nos. 6 and 7, the world's first ABWRs (Advanced Boiling Water Reactor), started commercial operation on November 7, 1996 and July 2, 1997, respectively, and continued their commercial operation with a high capacity factor, low occupational radiation exposure and radioactive waste. Units 6 and 7 were in their 3rd cycle operation until 25th April 1999 and 1st November 1999, respectively. Thermal efficiency was 35.4-35.8% (design thermal efficiency: 34.5%) during these period, demonstrating better performance than that of BWR-5 (design thermal efficiency: 33.4%). Nos. 6 and 7 have experienced 2 annual outages. The first outage of unit No. 6 started on November 20, 1997 and was completed within 61 days (including 6 New Year holidays), and the second outage started on March 13, 1999 and was completed within 44 days. The first annual outage of unit No. 7 started on May 27, 1998, earlier than it would normally have been, to avoid an annual outage during the summer, and was completed within 55 days, and the second outage started on September 18th, 1999 and was completed within 45 days, All annual outages were carried out within a very short time period without any severe malfunctions, including newly designed ABWR systems and equipment. As the first outage in Japan, 55 days is a very short period, despite the fact that the Nos. 6 and 7 are the first ABWRs in the world and the largest capacity units in Japan. The total occupational radiation exposure of No. 6 was 300 man-mSv (1st outage) and 331 man-mSv (2nd outage). That of Unit 7 was 153 man-mSv (1st outage) Those of unit No. 6 were at the same level as those of unit No. 3, which is the latest design 1100MW(e) BWR-5. That of unit No. 7 was the lowest ever at Kashiwazaki-Kariwa nuclear power station. The drums of radioactive waste discharged during the annual outage numbered 54 (1st outage) for No. 6 and 62 (1st outage) for No. 7, which was less than the design target of 100

Hitachi's experience and achievements in ABWR construction

International Nuclear Information System (INIS)

Inoue, Takashi; Miura, Jun; Murayama, Kouichi

2003-01-01

Hitachi, Ltd. (Hitachi) is taking charge of the turbine island of Chubu Electric Power Company's Hamaoka nuclear power plant Unit No. 5 (Hamaoka 5), and all facilities of Hokuriku Electric Power Company's Shika Nuclear Power Station Unit No. 2 (Shika 2). Hitachi has achieved shorter schedules and lower costs in the two projects with the following construction methods. First, Hitachi applied modularization and the area-by-area method. These methods enable us to undertake building and installation work in parallel. Second, the floor packaging construction method enables us to finish installation in every area earlier than conventional methods. Moreover, manpower can be distributed evenly. This paper reports the application of these advanced construction methods to the Hamaoka 5 and Shika 2 projects. Installations such as RCCV modules and condenser blocks are presented in this paper. In addition, Hitachi is using 3D-CAD engineering and our own construction management system to optimize the planning of construction procedures and the installation of equipment in these projects. Thus Hitachi is always striving to improve the reliability, safety and economy of ABWR construction technologies. (author)

Defence-in-depth concept for the EU-ABWR

Energy Technology Data Exchange (ETDEWEB)

Yamazaki, Hiroshi; Fuchs, Steffen; Takada, Toshiaki; Kataoka, Kazuyoshi [Toshiba International Limited (Japan)

2013-07-01

The current defence-in-depth (DiD) concept has been established by the Reactor Harmonization Working Group (RHWG) of Western European Nuclear Regulators Association (WENRA). Principally the DiD concept was already part of the very early power reactor designs. However, additional considerations have been done in order to take plant conditions into account which are beyond the original design basis. The most recent advancements have been done based on major lessons learned from the Fukushima Dai-Ichi accident. Especially for new nuclear reactors it has to be demonstrated that DiD aspects have been considered in their design. Currently Toshiba is adapting its Advanced Boiling Water Reactor (ABWR) for the European market, at first in Finland. This presentation aims to describe how the new DiD concept has been applied to achieve the safety goals for a modern reactor type and to ensure a design that can be licensed in Western Europe. (orig.)

Development of advanced concept for shortening construction period of ABWR plant (part 2)

International Nuclear Information System (INIS)

Fujita, Tomohiro; Satoh, Takashi

2003-01-01

The construction of the first building fully applying SC structure, which is indispensable for shortening construction period of ABWR plants (21.5 months from the first concrete work to fuel loading), has been started since August 2002 in Japan. Before the construction start, a pre-construction test with some actual size SC panels was carried out to confirm the SC modular construction method. The outline of the design and the construction of the first full-SC building and the results of the pre-construction test are reported. (author)

Domestic and overseas development of advanced boiling water reactors

International Nuclear Information System (INIS)

Hatazawa, Mamoru; Fuchino, Satoshi; Nakada, Kotaro

2010-01-01

Since Toshiba delivered the world's first advanced boiling water reactor (ABWR) to The Tokyo Electric Power Company, Inc. in 1996, we have been devoting continuous efforts to the construction and operational support of ABWR systems as major products. We are now promoting the construction of domestic and overseas ABWR systems along with the standardization of ABWRs. We are also engaged in the research and development of core technologies to support further promotion of ABWRs as a concurrent solution to the issues of global warming and energy security for individual countries. (author)

Development of Advanced Concept for Shortening Construction Period of ABWR Plant

International Nuclear Information System (INIS)

Hiroshi Ijichi; Toshio Yamashita; Masahiro Tsutagawa; Hiroya Mori; Nobuaki Ooshima; Jun Miura; Minoru Kanechika; Nobuaki Miura

2002-01-01

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

Heat removal capability of core-catcher with inclined cooling channels

International Nuclear Information System (INIS)

Suzuki, Y.; Tahara, M.; Kurita, T.; Hamazaki, R.; Morooka, S.

2009-01-01

A core-catcher is one of the mitigation systems that provide functions of molten corium cooling and stabilization during a severe accident. Toshiba has been developing a compact core-catcher to be placed at the lower drywell floor in the containment vessel for the next generation BWR as well as near term ABWR. This paper presents the evaluation of heat removal capability of the core-catcher with inclined cooling channels, our verification status and plan. The heat removal capability of the core-catcher is analyzed by using the newly developed two-phase flow analysis code which incorporates drift flux parameters for inclined channels and the CHF correlation obtained from SULTAN tests. Effects of geometrical parameters such as the inclination and the gap size of the cooling channel on the heat removal capability are also evaluated. These results show that the core-catcher has sufficient capability to cool the molten corium during a severe accident. Based on the analysis, it has been shown that the core-catcher has an efficient capability of heat removal to cool the molten corium. (author)

Participation in the ABWR Man-Machine interface design. Applicability to the Spanish Electrical Sector

International Nuclear Information System (INIS)

Rodriguez, C.; Manrique Martin, A.; Nunez, J.

1997-01-01

Project coordinated by DTN within the advanced reactor programme. Participation in the design activities for the Advanced Boiling Water Reactor (ABWR) man-machine interface was divided into two phases: Phase I: Preparation of drawings for designing, developing and assessing the advanced control room Phase II: Application of these drawings in design activities Participation in this programme has led to the following possible future applications to the electrical sector: 1. Design and implementation of man-machine interfaces 2. Human factor criteria 3. Assessment of man-machine interfaces 4. Functional specification, computerised operating procedures 5. Computerised alarm prototypes. (Author)

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

Study of Pu consumption in Advanced Light Water Reactors

International Nuclear Information System (INIS)

1993-01-01

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology

Sodium hydroxide injection passivation work for the reactor water clean-up system in a new ABWR plant

International Nuclear Information System (INIS)

Wen, Tung-Jen; Lu, Ju-Huang

2012-09-01

Several studies have identified that Co-58 and Co-60 as the primary source of radiation build up on out-of-core components in new BWR plants. The deposition rate of Co on stainless steel and carbon steel is shown to be controlled mainly by the thickness of oxide films and its morphology formed through pretreatment. The passivation treatment was implemented accordingly at Lungmen unit 1 in an ABWR plant in September 2010. It is determined that the passivation conditions should be maintained at the temperature of 180∼230 deg. C, pH of 8.0∼8.5 and dissolved oxygen content over 400 ppb. The films would provide effective protection against radioactive deposition. The application of the pre-filming process on piping before the pre-operation is done during the flow induced vibration test (FIV) period. The protectiveness of stable magnetite can be increased by the pH control under the specific condition. The pre-filming control process and evaluation of passivation effectiveness is discussed in detail based on the surface analysis of the passivated specimens. Many efforts have been devoted to sodium hydroxide injection method for pH control of the system through the filter demineralizer under smooth operation. A comparison of test specimens on the properties of oxide film formed between laboratory and in-plant tests through alkaline treatment are also shown in this report. (authors)

Evaluating nuclear physics inputs in core-collapse supernova models

Science.gov (United States)

Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

SCORPIO-BWR: status and future plans

International Nuclear Information System (INIS)

Porsmyr, Jan; Bodal, Terje; Beere, William H.

2004-01-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

SCORPIO-BWR: status and future plans

Energy Technology Data Exchange (ETDEWEB)

Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

2004-07-01

Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO

Initiatives to reduce the occupational radiation exposure of ABWR plants

International Nuclear Information System (INIS)

Hirasawa, Hajime; Urata, Hidehiro; Ueda, Taku; Yamamoto, Seiji; Yaita, Yumi

2014-01-01

Toshiba has carried out radiation exposure reduction by radiation level reduction, as reduction of reactor water activated corrosion products concentration, reduction of activated corrosion products deposition and radiation shielding, and exposure time reduction, as remote control and improvement of maintenance work procedures. Water chemistry has been mainly carried out reduction of reactor water activated corrosion products concentration and reduction of activated corrosion products deposition in radiation level reduction. The reduction measures of reactor water activated corrosion products concentration are mainly reduction of iron crud concentration and reduction of cobalt ion concentration. The activated corrosion products deposition are reduced by the means of water quality control and the surface treatment. Water quality control for reduction of activated corrosion products deposition moves to ultra low iron high nickel control from Ni/Fe ratio control. The surface treatments are adopted to the stainless steel piping and carbon steel piping. As a measure further to radiation exposure reduction for ABWR (Advanced Boiling Water Reactors), we report on the effect of generation amount reduction by the adoption of alternate material and the effect of deposition reduction by material change of piping and the adoption of advanced water quality control, etc. (author)

HTR core physics and transient analyses by the Panthermix code system

International Nuclear Information System (INIS)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

2005-01-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

CHEMICAL AND PHYSICAL CHARACTERIZATION OF COLLAPSING LOW-MASS PRESTELLAR DENSE CORES

Energy Technology Data Exchange (ETDEWEB)

Hincelin, U. [Department of Chemistry, University of Virginia, Charlottesville, VA 22904 (United States); Commerçon, B. [Ecole Normale Supérieure de Lyon, CRAL, UMR 5574 du CNRS, Université Lyon I, 46 Allée d’Italie, F-69364 Lyon cedex 07 (France); Wakelam, V.; Hersant, F.; Guilloteau, S. [Univ. Bordeaux, LAB, UMR 5804, F-33270, Floirac (France); Herbst, E., E-mail: ugo.hincelin@gmail.com [Departments of Chemistry and Astronomy, University of Virginia, Charlottesville, VA 22904 (United States)

2016-05-01

The first hydrostatic core, also called the first Larson core, is one of the first steps in low-mass star formation as predicted by theory. With recent and future high-performance telescopes, the details of these first phases are becoming accessible, and observations may confirm theory and even present new challenges for theoreticians. In this context, from a theoretical point of view, we study the chemical and physical evolution of the collapse of prestellar cores until the formation of the first Larson core, in order to better characterize this early phase in the star formation process. We couple a state-of-the-art hydrodynamical model with full gas-grain chemistry, using different assumptions for the magnetic field strength and orientation. We extract the different components of each collapsing core (i.e., the central core, the outflow, the disk, the pseudodisk, and the envelope) to highlight their specific physical and chemical characteristics. Each component often presents a specific physical history, as well as a specific chemical evolution. From some species, the components can clearly be differentiated. The different core models can also be chemically differentiated. Our simulation suggests that some chemical species act as tracers of the different components of a collapsing prestellar dense core, and as tracers of the magnetic field characteristics of the core. From this result, we pinpoint promising key chemical species to be observed.

Needs of nuclear data for advanced light water reactor

International Nuclear Information System (INIS)

Chaki, Masao

2008-01-01

Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO 2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO 2 cores and the MOX cores. (author)

The online simulation of core physics in nuclear power plant

International Nuclear Information System (INIS)

Zhao Qiang

2005-01-01

The three-dimensional power distribution in core is one of the most important status variables of nuclear reactor. In order to monitor the 3-D in core power distribution timely and accurately, the online simulation system of core physics was designed in the paper. This system combines core physics simulation with the data, which is from the plant and reactor instrumentation. The design of the system consists of the hardware part and the software part. The online simulation system consists of a main simulation computer and a simulation operation station. The online simulation system software includes of the real-time simulation support software, the system communication software, the simulation program and the simulation interface software. Two-group and three-dimensional neutron kinetics model with six groups delayed neutrons was used in the real-time simulation of nuclear reactor core physics. According to the characteristics of the nuclear reactor, the core was divided into many nodes. Resolving the neutron equation, the method of separate variables was used. The input data from the plant and reactor instrumentation system consist of core thermal power, loop temperatures and pressure, control rod positions, boron concentration, core exit thermocouple data, Excore detector signals, in core flux detectors signals. There are two purposes using the data, one is to ensure that the model is as close as the current actual reactor condition, and the other is to calibrate the calculated power distribution. In this paper, the scheme of the online simulation system was introduced. Under the real-time simulation support system, the simulation program is being compiled. Compared with the actual operational data, the elementary simulation results were reasonable and correct. (author)

HTR core physics and transient analyses by the Panthermix code system

Energy Technology Data Exchange (ETDEWEB)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

Normal Mode Derived Models of the Physical Properties of Earth's Outer Core

Science.gov (United States)

Irving, J. C. E.; Cottaar, S.; Lekic, V.; Wu, W.

2017-12-01

Earth's outer core, the largest reservoir of metal in our planet, is comprised of an iron alloy of an uncertain composition. Its dynamical behaviour is responsible for the generation of Earth's magnetic field, with convection driven both by thermal and chemical buoyancy fluxes. Existing models of the seismic velocity and density of the outer core exhibit some variation, and there are only a small number of models which aim to represent the outer core's density.It is therefore important that we develop a better understanding of the physical properties of the outer core. Though most of the outer core is likely to be well mixed, it is possible that the uppermost outer core is stably stratified: it may be enriched in light elements released during the growth of the solid, iron enriched, inner core; by elements dissolved from the mantle into the outer core; or by exsolution of compounds previously dissolved in the liquid metal which will eventually be swept into the mantle. The stratified layer may host MAC or Rossby waves and it could impede communication between the chemically differentiated mantle and outer core, including screening out some of the geodynamo's signal. We use normal mode center frequencies to estimate the physical properties of the outer core in a Bayesian framework. We estimate the mineral physical parameters needed to best produce velocity and density models of the outer core which are consistent with the normal mode observations. We require that our models satisfy realistic physical constraints. We create models of the outer core with and without a distinct uppermost layer and assess the importance of this region.Our normal mode-derived models are compared with observations of body waves which travel through the outer core. In particular, we consider SmKS waves which are especially sensitive to the uppermost outer core and are therefore an important way to understand the robustness of our models.

Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun

2001-08-01

This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given.

Study on Reactor Physics Characteristic of the PWR Core Using UO2

International Nuclear Information System (INIS)

Tukiran Surbakti

2009-01-01

Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

Advanced boiling water reactor

International Nuclear Information System (INIS)

Nishimura, N.; Nakai, H.; Ross, M.A.

1999-01-01

In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

Energy Technology Data Exchange (ETDEWEB)

Forget, Benoit; Smith, Kord; Kumar, Shikhar; Rathbun, Miriam; Liang, Jingang

2018-04-11

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevant multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.

Physical properties of sidewall cores from Decatur, Illinois

Science.gov (United States)

Morrow, Carolyn A.; Kaven, Joern; Moore, Diane E.; Lockner, David A.

2017-10-18

To better assess the reservoir conditions influencing the induced seismicity hazard near a carbon dioxide sequestration demonstration site in Decatur, Ill., core samples from three deep drill holes were tested to determine a suite of physical properties including bulk density, porosity, permeability, Young’s modulus, Poisson’s ratio, and failure strength. Representative samples of the shale cap rock, the sandstone reservoir, and the Precambrian basement were selected for comparison. Physical properties were strongly dependent on lithology. Bulk density was inversely related to porosity, with the cap rock and basement samples being both least porous (

Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

2007-07-01

Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

Methodology for reactor core physics analysis - part 2

International Nuclear Information System (INIS)

Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

1992-12-01

The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

Micro-Reactor Physics of MOX-Fueled Core

International Nuclear Information System (INIS)

Takeda, T.

2001-01-01

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

The Six Core Theories of Modern Physics

Science.gov (United States)

Stevens, Charles F.

1996-09-01

Charles Stevens, a prominent neurobiologist who originally trained as a biophysicist (with George Uhlenbeck and Mark Kac), wrote this book almost by accident. Each summer he found himself reviewing key areas of physics that he had once known and understood well, for use in his present biological research. Since there was no book, he created his own set of notes, which formed the basis for this brief, clear, and self-contained summary of the basic theoretical structures of classical mechanics, electricity and magnetism, quantum mechanics, statistical physics, special relativity, and quantum field theory. The Six Core Theories of Modern Physics can be used by advanced undergraduates or beginning graduate students as a supplement to the standard texts or for an uncluttered, succinct review of the key areas. Professionals in such quantitative sciences as chemistry, engineering, computer science, applied mathematics, and biophysics who need to brush up on the essentials of a particular area will find most of the required background material, including the mathematics.

Integrating the English Language Arts Common Core State Standards into Physical Education

Science.gov (United States)

James, Alisa R.; Bullock, Kerri

2015-01-01

Physical education teachers are expected to implement the English language arts (ELA) Common Core State Standards (CCSS) in their instruction. This has proved to be challenging for many physical educators. The purpose of this article is to provide developmentally appropriate examples of how to incorporate the ELA CCSS into physical education,…

Development of spectral history methods for pin-by-pin core analysis method using three-dimensional direct response matrix

International Nuclear Information System (INIS)

Mitsuyasu, T.; Ishii, K.; Hino, T.; Aoyama, M.

2009-01-01

Spectral history methods for pin-by-pin core analysis method using the three-dimensional direct response matrix have been developed. The direct response matrix is formalized by four sub-response matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the burn-up effect related to spectral history. One of the methods is to evaluate the nodal burn-up spectrum obtained using the out-going neutron current. The other is to correct the fuel rod neutron production rates obtained the pin-by-pin correction. These spectral history methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error can be reduced by half during burn-up, the nodal neutron production rates errors can be reduced by 30% or more. The root-mean-square differences between the relative fuel rod neutron production rate distributions can be reduced within 1.1% error. This means that these methods can accurately reflect the effects of intra- and inter-assembly heterogeneities during burn-up and can be used for core analysis. Core analysis with the DRM method was carried out for an ABWR quarter core and it was found that both thermal power and coolant-flow distributions were smoothly converged. (authors)

Fast breeder physics and nuclear core design

International Nuclear Information System (INIS)

Marth, W.; Schroeder, R.

1983-07-01

This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

Toward a mineral physics reference model for the Moon's core.

Science.gov (United States)

Antonangeli, Daniele; Morard, Guillaume; Schmerr, Nicholas C; Komabayashi, Tetsuya; Krisch, Michael; Fiquet, Guillaume; Fei, Yingwei

2015-03-31

The physical properties of iron (Fe) at high pressure and high temperature are crucial for understanding the chemical composition, evolution, and dynamics of planetary interiors. Indeed, the inner structures of the telluric planets all share a similar layered nature: a central metallic core composed mostly of iron, surrounded by a silicate mantle, and a thin, chemically differentiated crust. To date, most studies of iron have focused on the hexagonal closed packed (hcp, or ε) phase, as ε-Fe is likely stable across the pressure and temperature conditions of Earth's core. However, at the more moderate pressures characteristic of the cores of smaller planetary bodies, such as the Moon, Mercury, or Mars, iron takes on a face-centered cubic (fcc, or γ) structure. Here we present compressional and shear wave sound velocity and density measurements of γ-Fe at high pressures and high temperatures, which are needed to develop accurate seismic models of planetary interiors. Our results indicate that the seismic velocities proposed for the Moon's inner core by a recent reanalysis of Apollo seismic data are well below those of γ-Fe. Our dataset thus provides strong constraints to seismic models of the lunar core and cores of small telluric planets. This allows us to propose a direct compositional and velocity model for the Moon's core.

ABWR2. Innovative passive containment cooling system for the innovative ABWR

International Nuclear Information System (INIS)

Sato, Takashi; Matsumoto, Keiji

2015-01-01

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. The iB1350 uses innovative passive containment cooling system (iPCCS). The iPCCS is a part of the in-containment filtered venting system (IFVS). The vent pipe is submerged in the IFVS tank in the outer well (OW) of the Mark W containment. The conventional PCCS has a suction pipe only from the dry well (DW). On the contrary, the iPCCS has two suction pipes. One is normally opened to the wet well (WW) and another normally closed to the DW. The suction pipe in the conventional design cannot be connected to the WW because the PCCS vent pipe is connected to the WW. A PCCS functions using differential pressure between two nodes to discharge noncondensable gases in a PCCS heat exchanger (Hx). A suction pipe and a vent pipe must be connected to different nodes to use differential pressure. Therefore, the conventional PCCS never can cool the S/P. Although the S/P is the in-containment heat sink, heat up of the S/P is the most unfavorable for the conventional PCCS. In order to use the PCCS the conventional design must discharge steam directly into the DW instead of the S/P. Therefore, the conventional PCCS must open depressurization valves (DPV) at a SBO if the isolation condenser (IC) fails. On the contrary, the iPCCS can cool the S/P directly using the suction pipe connected to the WW and without DPV. Instead of DPV the iB1350 has modulating valves (MV) of which discharge lines are submerged in the S/P. Even if the IC fails at a SBO, the iB1350 can cool the core using the severe accident feedwater system (SAFWS), the SRV or the MV, and the iPCCS. The SAFWS makes up the core. The decay heat is carried by steam to the S/P through the SRV or the MV. The S/P works as in-containment heat sink. Once the S/P starts boiling the iPCCS automatically initiates cooling of the steam from the S/P. In the case of a core melt accident, a certain amount of FP is released into the S/P and heats up the S/P. Once the S

Analysis of high moderation full MOX BWR core physics experiments BASALA

International Nuclear Information System (INIS)

Ishii, Kazuya; Ando, Yoshihira; Takada, Naoyuki; Kan, Taro; Sasagawa, Masaru; Kikuchi, Tsukasa; Yamamoto, Toru; Kanda, Ryoji; Umano, Takuya

2005-01-01

Nuclear Power Engineering Corporation (NUPEC) has performed conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program BASALA following MISTRAL. They were devoted to measuring the core physics parameters of such advanced cores. The MISTRAL program consists of one reference UO 2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. As for MISTRAL, the analysis results have already been reported on April 2003. The BASALA program consists of two high moderation full MOX BWR mock-up cores for operating and cold stand-by conditions. NUPEC has analyzed the experimental results of BASALA with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. The analysis results of MISTRAL and BASALA indicate that these applied analysis methods have the same accuracy for the UO 2 and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores. (author)

Basic philosophy of the safety design of the Toshiba boiling water reactor

International Nuclear Information System (INIS)

Sato, T.

1992-01-01

This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

Physical properties of a sediment core from the Central Indian Basin

Digital Repository Service at National Institute of Oceanography (India)

Khadge, N.H.

A box core of 7.5 m was collected from the Central Indian Basin for the purpose of geotechnical studies and depthwise variation of physical properties and clay mineralogy. Water content, Atterberg limits, specific gravity are measured at regular...

The reactor core configuration and important systems related to physics tests of Daya Bay NPP

International Nuclear Information System (INIS)

Tao Shaoping

1995-06-01

A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (RIC), the out-of-core nuclear instrumentation system (RPN), and the LOCA surveillance system (LSS), the centralized data processing system (KIT) and the test data acquisition system (KDO). In addition, that the adjustment and evaluation of boron concentration related to other systems, for example the reactor coolant system (RCP), the chemical and volume control system (RCV), the reactor boron and water makeup system (REA), the nuclear sampling system (REN) and the reactor control system (RRC), etc. is also described. Analysis of these systems helps not only to familiarize their functions and acquires a deepen understanding for the principle procedure, points for attention and technical key of the core physics tests, but also to further analyze the test results. (3 refs., 11 figs., 1 tab.)

Core journals that publish clinical trials of physical therapy interventions.

Science.gov (United States)

Costa, Leonardo Oliveira Pena; Moseley, Anne M; Sherrington, Catherine; Maher, Christopher G; Herbert, Robert D; Elkins, Mark R

2010-11-01

The objective of this study was to identify core journals in physical therapy by identifying those that publish the most randomized controlled trials of physical therapy interventions, provide the highest-quality reports of randomized controlled trials, and have the highest journal impact factors. This study was an audit of a bibliographic database. All trials indexed in the Physiotherapy Evidence Database (PEDro) were analyzed. Journals that had published at least 80 trials were selected. The journals were ranked in 4 ways: number of trials published; mean total PEDro score of the trials published in the journal, regardless of publication year; mean total PEDro score of the trials published in the journal from 2000 to 2009; and 2008 journal impact factor. The top 5 core journals in physical therapy, ranked by the total number of trials published, were Archives of Physical Medicine and Rehabilitation, Clinical Rehabilitation, Spine, British Medical Journal (BMJ), and Chest. When the mean total PEDro score was used as the ranking criterion, the top 5 journals were Journal of Physiotherapy, Journal of the American Medical Association (JAMA), Stroke, Spine, and Clinical Rehabilitation. When the mean total PEDro score of the trials published from 2000 to 2009 was used as the ranking criterion, the top 5 journals were Journal of Physiotherapy, JAMA, Lancet, BMJ, and Pain. The most highly ranked physical therapy-specific journals were Physical Therapy (ranked eighth on the basis of the number of trials published) and Journal of Physiotherapy (ranked first on the basis of the quality of trials). Finally, when the 2008 impact factor was used for ranking, the top 5 journals were JAMA, Lancet, BMJ, American Journal of Respiratory and Critical Care Medicine, and Thorax. There were no significant relationships among the rankings on the basis of trial quality, number of trials, or journal impact factor. Physical therapists who are trying to keep up-to-date by reading the best

Fast reactors fuel cycle core physics results from the CAPRA-CADRA programme

Energy Technology Data Exchange (ETDEWEB)

Vasile, A.; Rimpault, G.; Tommasi, J.; Saint Jean, C. de; Delpech, M. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Hesketh, K. [BNFL, Inc., Denver, CO (United States); Beaumont, H.M.; Sunderland, R.E. [NNC Ltd. (United Kingdom); Newton, T.; Smith, P. [AEA Technology (United Kingdom); Raedt, Ch. de [SCK.CEN, Mol (Belgium); Vambenepe, G. [Electricite de France (EDF), 75 - Paris (France); Lefevre, J.C. [FRAMATOME, 92 - Paris-La-Defence (France); Maschek, W.; Haas, D

2001-07-01

This paper presents an overview of fast reactor core physics results obtained in the context of the CAPRA-CADRA European collaborative programme, whose aim is to investigate a broad range of possible options for plutonium and radioactive waste management. Different types of fast reactors have been studied to evaluate their potential capabilities with respect to the long term management of plutonium, minor actinides (MAs) and long- lived fission products (LLFPs). Among the several options aiming at reducing waste and consequently radio toxicity are: homogeneous recycling of Minor Actinides, heterogeneous recycling of Minor Actinides either without or with moderation, dedicated critical cores (fuelled mainly with Minor Actinides) and Accelerator Driven System (ADS) variants. In order to achieve a detailed understanding of the potential of the various options, advanced core physics methods have been implemented and tested and applied, for example, to improving control rod modeling and to studying safety aspects. There has also been code development and experimental work carried out to improve the understanding of fuel performance behaviors. (author)

Fast reactors fuel cycle core physics results from the CAPRA-CADRA programme

International Nuclear Information System (INIS)

Vasile, A.; Rimpault, G.; Tommasi, J.; Saint Jean, C. de; Delpech, M.; Hesketh, K.; Beaumont, H.M.; Sunderland, R.E.; Newton, T.; Smith, P.; Raedt, Ch. de; Vambenepe, G.; Lefevre, J.C.; Maschek, W.; Haas, D

2001-01-01

This paper presents an overview of fast reactor core physics results obtained in the context of the CAPRA-CADRA European collaborative programme, whose aim is to investigate a broad range of possible options for plutonium and radioactive waste management. Different types of fast reactors have been studied to evaluate their potential capabilities with respect to the long term management of plutonium, minor actinides (MAs) and long- lived fission products (LLFPs). Among the several options aiming at reducing waste and consequently radio toxicity are: homogeneous recycling of Minor Actinides, heterogeneous recycling of Minor Actinides either without or with moderation, dedicated critical cores (fuelled mainly with Minor Actinides) and Accelerator Driven System (ADS) variants. In order to achieve a detailed understanding of the potential of the various options, advanced core physics methods have been implemented and tested and applied, for example, to improving control rod modeling and to studying safety aspects. There has also been code development and experimental work carried out to improve the understanding of fuel performance behaviors. (author)

Core physics calculation and analysis for SNRE

International Nuclear Information System (INIS)

Xie Jiachun; Zhao Shouzhi; Jia Baoshan

2010-01-01

Five different precise calculation models have been set up for Small Nuclear Rocket Engine (SNRE) core based on MCNP code, and then the effective multiplication constant, drum control worth and power distribution were calculated. The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation, but a detailed description of elements have to be used in the element internal power distribution calculation. The results of physics parameters show that the basic characteristics of SNRE are reasonable. The drum control worth is sufficient. The power distribution is symmetrical and reasonable. All of the parameters can satisfy the design requirement. (authors)

Nursing physical assessment for patient safety in general wards: reaching consensus on core skills.

Science.gov (United States)

Douglas, Clint; Booker, Catriona; Fox, Robyn; Windsor, Carol; Osborne, Sonya; Gardner, Glenn

2016-07-01

To determine consensus across acute care specialty areas on core physical assessment skills necessary for early recognition of changes in patient status in general wards. Current approaches to physical assessment are inconsistent and have not evolved to meet increased patient and system demands. New models of nursing assessment are needed in general wards that ensure a proactive and patient safety approach. A modified Delphi study. Focus group interviews with 150 acute care registered nurses at a large tertiary referral hospital generated a framework of core skills that were developed into a web-based survey. We then sought consensus with a panel of 35 senior acute care registered nurses following a classical Delphi approach over three rounds. Consensus was predefined as at least 80% agreement for each skill across specialty areas. Content analysis of focus group transcripts identified 40 discrete core physical assessment skills. In the Delphi rounds, 16 of these were consensus validated as core skills and were conceptually aligned with the primary survey: (Airway) Assess airway patency; (Breathing) Measure respiratory rate, Evaluate work of breathing, Measure oxygen saturation; (Circulation) Palpate pulse rate and rhythm, Measure blood pressure by auscultation, Assess urine output; (Disability) Assess level of consciousness, Evaluate speech, Assess for pain; (Exposure) Measure body temperature, Inspect skin integrity, Inspect and palpate skin for signs of pressure injury, Observe any wounds, dressings, drains and invasive lines, Observe ability to transfer and mobilise, Assess bowel movements. Among a large and diverse group of experienced acute care registered nurses consensus was achieved on a structured core physical assessment to detect early changes in patient status. Although further research is needed to refine the model, clinical application should promote systematic assessment and clinical reasoning at the bedside. © 2016 John Wiley & Sons Ltd.

A Multi-Physics simulation of the Reactor Core using CUPID/MASTER

International Nuclear Information System (INIS)

Lee, Jae Ryong; Cho, Hyoung Kyu; Yoon, Han Young; Cho, Jin Young; Jeong, Jae Jun

2011-01-01

KAERI has been developing a component-scale thermal hydraulics code, CUPID. The aim of the code is for multi-dimensional, multi-physics and multi-scale thermal hydraulics analysis. In our previous papers, the CUPID code has proved to be able to reproduce multidimensional thermal hydraulic analysis by validated with various conceptual problems and experimental data. For the numerical closure, it adopts a three dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer. For the multi-scale analysis, the CUPID is on progress to merge into system-scale thermal hydraulic code, MARS. In the present paper, a multi-physics simulation was performed by coupling the CUPID with three dimensional neutron kinetics code, MASTER. The MASTER is merged into the CUPID as a dynamic link library (DLL). The APR1400 reactor core during control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results

Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

International Nuclear Information System (INIS)

Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

1994-01-01

After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

International Nuclear Information System (INIS)

Lell, R.M.; Hanan, N.A.

1987-01-01

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

International Nuclear Information System (INIS)

Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

2011-01-01

Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

Demountable toroidal fusion core facility for physics optimization and fusion engineering

International Nuclear Information System (INIS)

Bogart, S.L.; Wagner, C.E.; Krall, N.A.; Dalessandro, J.A.; Weggel, C.F.; Lund, K.O.; Sedehi, S.

1986-01-01

Following a successful compact ignition tokamak (CIT) experiment, a fusion facility will be required for physics optimization (POF) and fusion engineering research (FERF). The POF will address issues such as high-beta operation, current drive, impurity control, and will test geometric and configurational variations such as the spherical torus or the reversed-field pinch (RFP). The FERF will be designed to accumulate rapidly a large neutron dose in prototypical fusion subsystems exposed to radiation. Both facilities will require low-cost replacement cores and rapid replacement times. The Demountable Toroidal Fusion Core (DTFC) facility is designed to fulfill these requirements. It would be a cost-effective stepping stone between the CIT and a demonstration fusion reactor

A critical experimental study of integral physics parameters in simulated LMFBR meltdown cores

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Wade, D.C.; Bucher, R.G.; Smith, D.M.; McKnight, R.D.; Lesage, L.G.

1978-01-01

Integral physics parameters of several representative, idealized meltdown LMFBR configurations were measured in mockup critical assemblies on the ZPR-9 reactor at Argonne National Laboratory. The experiments were designed to provide data for the validation of analytical methods used in the neutronics part of LMFBR accident analysis. Large core distortions were introduced in these experiments (involving 18.5% core volume) and the reactivity worths of configuration changes were determined. The neutronics parameters measured in the various configurations showed large changes upon core distortion. Both diffusion theory and transport theory methods were shown to mispredict the experimental configuration eigenvalues. In addition, diffusion theory methods were shown to result in a non-conservative misprediction of the experimental configuration change worths. (author)

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

International Nuclear Information System (INIS)

Lee, J.H.; Oka, Y.; Koshizuka, S.

1999-01-01

The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics

Overview of JSPS Core-to-Core Program: Forming Research and Educational Hubs of Medical Physics.

Science.gov (United States)

Koizumi, Masahiko; Takashina, Masaaki

To foster medical physicists, we introduce the achievement we made since 2011 under the national research project of the Japan Society for the Promotion of Science (JSPS) Core-to-Core program; 'Forming Research and Educational Hubs of Medical Physics.' On this basis and under the JSPS program, we promoted research and educational exchange with Indiana University (IU) in USA, University of Groningen (The UG) in the Netherland and other cooperating institutions such as University of Minnesota (UM).A total of 23 students and researchers were sent. UG accepted the most among three institutions. In turn, 12 foreign researchers including post-doctor fellows came to Japan for academic seminars or educational lectures.Fifteen international seminars were held; 8 in Japan, 4 in USA, and 3 in the Netherland.Lots of achievement were made through these activities in 5 years. Total of 23 research topics at the international conferences were presented. Total of 12 articles were published in international journals.This program clearly promoted the establishment of international collaboration, and many young researchers and graduate students were exchanged and collaborated with foreign researchers.

Verification of MVP-II and SRAC2006 code to the core physics vera benchmark problem

International Nuclear Information System (INIS)

Jati Susilo

2014-01-01

In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT) of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by. Westinghouse, arranged from 193 unit of 17 x 17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC) and hot zero power (HZP). In this calculation, MVP-II and CITATION module of SRAC2006 computer code has been used with ENDF/B-VII.0. cross section data library. Calculation result showed that differences value of k-eff for the core at controlled and uncontrolled condition between reference with MVP-II (-0,07% and -0,014%) and SRAC2006 (0,92% and 0,99%) are very small or below 1%. Differences value of radial power peaking factor at controlled and uncontrolled of the core between reference value with MVP-II are 0,38% and 1,53%, even though with SRAC2006 are 1,13% and -2,45%. It can be said that the calculation result by both computer code showing suitability with reference value. In order to determinate of criticality of the core, the calculation result using MVP-II code is more conservative compare with SRAC2006 code. (author)

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

Energy Technology Data Exchange (ETDEWEB)

Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Blaise, P. [CEA, DEN, DER, SPEX Experimental Programs Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

2015-07-01

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

Monte Carlo simulation of core physics parameters of the Syrian MNSR reactor

International Nuclear Information System (INIS)

Khattab, K.; Sulieman, I.

2011-01-01

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis using the MCNP-4C code. The continuous energy neutron cross sections were evaluated from the ENDF/B-VI library. This model is used in this paper to calculate the following reactor core physics parameters: the clean cold core excess reactivity, calibration of the control rod and calculation its shut down margin, calibration of the top beryllium shim plate reflector, the axial neutron flux distributions in the inner and outer irradiation positions and calculations of the prompt neutron life time (ι p ) and the effective delayed neutron fraction ( β e ff). Good agreements are noticed between the calculated and the measured results. These agreements indicate that the established model is an accurate representation of Syrian MNSR core and will be used for other calculations in the future. (author)

THE BLAST SURVEY OF THE VELA MOLECULAR CLOUD: PHYSICAL PROPERTIES OF THE DENSE CORES IN VELA-D

International Nuclear Information System (INIS)

Olmi, Luca; Angles-Alcazar, Daniel; Ade, Peter A. R.; Griffin, Matthew; Hargrave, Peter C.; Bock, James J.; Chapin, Edward L.; Halpern, Mark; Marsden, Gaelen; De Luca, Massimo; Devlin, Mark J.; Dicker, Simon; Klein, Jeff; Elia, Davide; Fazio, Giovanni G.; Marengo, Massimo; Giannini, Teresa; Lorenzetti, Dario; Gundersen, Joshua O.; Hughes, David H.

2009-01-01

The Balloon-borne Large-Aperture Submillimeter Telescope (BLAST) carried out a 250, 350, and 500 μm survey of the galactic plane encompassing the Vela Molecular Ridge, with the primary goal of identifying the coldest dense cores possibly associated with the earliest stages of star formation. Here, we present the results from observations of the Vela-D region, covering about 4 deg 2 , in which we find 141 BLAST cores. We exploit existing data taken with the Spitzer MIPS, IRAC, and SEST-SIMBA instruments to constrain their (single-temperature) spectral energy distributions, assuming a dust emissivity index β = 2.0. This combination of data allows us to determine the temperature, luminosity, and mass of each BLAST core, and also enables us to separate starless from protostellar sources. We also analyze the effects that the uncertainties on the derived physical parameters of the individual sources have on the overall physical properties of starless and protostellar cores, and we find that there appear to be a smooth transition from the pre- to the protostellar phase. In particular, for protostellar cores we find a correlation between the MIPS24 flux, associated with the central protostar, and the temperature of the dust envelope. We also find that the core mass function of the Vela-D cores has a slope consistent with other similar (sub)millimeter surveys.

The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

Directory of Open Access Journals (Sweden)

LI U Ja

2017-02-01

Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations，the fi- nite element coupling in the transient electromagnetic field coupling，structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution，core magnetostrictive displacement，max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor，the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated，the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment，it is proved multi-physics coupling is an effective method for forecast of noise.

The American Society for Radiation Oncology's 2015 Core Physics Curriculum for Radiation Oncology Residents

International Nuclear Information System (INIS)

Burmeister, Jay; Chen, Zhe; Chetty, Indrin J.; Dieterich, Sonja; Doemer, Anthony; Dominello, Michael M.; Howell, Rebecca M.; McDermott, Patrick; Nalichowski, Adrian; Prisciandaro, Joann; Ritter, Tim; Smith, Chadd; Schreiber, Eric; Shafman, Timothy; Sutlief, Steven; Xiao, Ying

2016-01-01

Purpose: The American Society for Radiation Oncology (ASTRO) Physics Core Curriculum Subcommittee (PCCSC) has updated the recommended physics curriculum for radiation oncology resident education to improve consistency in teaching, intensity, and subject matter. Methods and Materials: The ASTRO PCCSC is composed of physicists and physicians involved in radiation oncology residency education. The PCCSC updated existing sections within the curriculum, created new sections, and attempted to provide additional clinical context to the curricular material through creation of practical clinical experiences. Finally, we reviewed the American Board of Radiology (ABR) blueprint of examination topics for correlation with this curriculum. Results: The new curriculum represents 56 hours of resident physics didactic education, including a 4-hour initial orientation. The committee recommends completion of this curriculum at least twice to assure both timely presentation of material and re-emphasis after clinical experience. In addition, practical clinical physics and treatment planning modules were created as a supplement to the didactic training. Major changes to the curriculum include addition of Fundamental Physics, Stereotactic Radiosurgery/Stereotactic Body Radiation Therapy, and Safety and Incidents sections, and elimination of the Radiopharmaceutical Physics and Dosimetry and Hyperthermia sections. Simulation and Treatment Verification and optional Research and Development in Radiation Oncology sections were also added. A feedback loop was established with the ABR to help assure that the physics component of the ABR radiation oncology initial certification examination remains consistent with this curriculum. Conclusions: The ASTRO physics core curriculum for radiation oncology residents has been updated in an effort to identify the most important physics topics for preparing residents for careers in radiation oncology, to reflect changes in technology and practice since

ASTRO's core physics curriculum for radiation oncology residents

International Nuclear Information System (INIS)

Klein, Eric E.; Balter, James M.; Chaney, Edward L.; Gerbi, Bruce J.; Hughes, Lesley

2004-01-01

In 2002, the Radiation Physics Committee of the American Society of Therapeutic Radiology and Oncology (ASTRO) appointed an Ad-hoc Committee on Physics Teaching to Medical Residents. The main initiative of the committee was to develop a core curriculum for physics education. Prior publications that have analyzed physics teaching have pointed to wide discrepancies among teaching programs. The committee was composed of physicists or physicians from various residency program based institutions. Simultaneously, members had associations with the American Association of Physicists in Medicine (AAPM), ASTRO, Association of Residents in Radiation Oncology (ARRO), American Board of Radiology (ABR), and the American College of Radiology (ACR). The latter two organizations' representatives were on the physics examination committees, as one of the main agendas was to provide a feedback loop between the examining organizations and ASTRO. The document resulted in a recommended 54-h course. Some of the subjects were based on American College of Graduate Medical Education (ACGME) requirements (particles, hyperthermia), whereas the majority of the subjects along with the appropriated hours per subject were devised and agreed upon by the committee. For each subject there are learning objectives and for each hour there is a detailed outline of material to be covered. Some of the required subjects/h are being taught in most institutions (i.e., Radiation Measurement and Calibration for 4 h), whereas some may be new subjects (4 h of Imaging for Radiation Oncology). The curriculum was completed and approved by the ASTRO Board in late 2003 and is slated for dissemination to the community in 2004. It is our hope that teaching physicists will adopt the recommended curriculum for their classes, and simultaneously that the ABR for its written physics examination and the ACR for its training examination will use the recommended curriculum as the basis for subject matter and depth of

78 FR 18375 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Science.gov (United States)

2013-03-26

... Pike, Rockville, Maryland. Thursday, April 11, 2013, Conference Room T2-B1, 11545 Rockville Pike..., ``Westinghouse BWR ECCS Evaluation Model: Supplement 5--Application to the ABWR,'' Revision 0 (Open/Closed)--The...-17116-P, ``Westinghouse BWR Emergency Core Coolant System (ECCS) Evaluation Model: Supplement 5,'' and...

Near-term and next-generation nuclear power plant concepts

International Nuclear Information System (INIS)

Shiga, Shigenori; Handa, Norihiko; Heki, Hideaki

2002-01-01

Near-term and next-generation nuclear reactors will be required to have high economic competitiveness in the deregulated electricity market, flexibility with respect to electricity demand and investment, and good public acceptability. For near-term reactors in the 2010s, Toshiba is developing an improved advanced boiling water reactor (ABWR) based on the present ABWR with newly rationalized systems and components; a construction period of 36 months, one year shorter than the current period; and a power lineup ranging from 800 MWe to 1,600 MWe. For future reactors in the 2020s and beyond, Toshiba is developing the ABWR-II for large-scale, centralized power sources; a supercritical water-cooled power reactor with high thermal efficiency for medium-scale power sources; a modular reactor with siting flexibility for small-scale power sources; and a small, fast neutron reactor with inherent safety for independent power sources. From the viewpoint of efficient uranium resource utilization, a low-moderation BWR core with a high conversion factor is also being developed. (author)

Verification of results of core physics on-line simulation by NGFM code

International Nuclear Information System (INIS)

Zhao Yu; Cao Xinrong; Zhao Qiang

2008-01-01

Nodal Green's Function Method program NGFM/TNGFM has been trans- planted to windows system. The 2-D and 3-D benchmarks have been checked by this program. And the program has been used to check the results of QINSHAN-II reactor simulation. It is proved that the NGFM/TNGFM program is applicable for reactor core physics on-line simulation system. (authors)

An assessment of coupling algorithms for nuclear reactor core physics simulations

Energy Technology Data Exchange (ETDEWEB)

Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

2016-04-15

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

International Nuclear Information System (INIS)

Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

2003-01-01

The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

Core physics design calculation of mini-type fast reactor based on Monte Carlo method

International Nuclear Information System (INIS)

He Keyu; Han Weishi

2007-01-01

An accurate physics calculation model has been set up for the mini-type sodium-cooled fast reactor (MFR) based on MCNP-4C code, then a detailed calculation of its critical physics characteristics, neutron flux distribution, power distribution and reactivity control has been carried out. The results indicate that the basic physics characteristics of MFR can satisfy the requirement and objectives of the core design. The power density and neutron flux distribution are symmetrical and reasonable. The control system is able to make a reliable reactivity balance efficiently and meets the request for long-playing operation. (authors)

Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-09-01

The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

Control room unfiltered in-leakage limit analysis of design-basis LOCA for Lungmen ABWR plant

International Nuclear Information System (INIS)

Tsai Chihming; Chang Chinjang; Yuann Yngruey

2014-01-01

In USNRC's Generic Letter 2003-01, 'Control Room Habitability,' it requests utilities provide information to demonstrate that the control room at each of their respective facilities complies with the current licensing and design bases, and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance. In particular, each utility is required to perform the control room in-leakage test to demonstrate that the unfiltered in-leakage rate is within that assumed in the licensing analyses. It must be ensured that the control room envelope habitability, in terms of radiation dose, is maintained during normal operations as well as design basis accidents. In view of this, a dose analysis has been performed to establish the control room unfiltered in-leakage limit which can be used as an acceptance criterion for the in-leakage test. The analysis in this study is for Lungmen ABWR plant. The plant has twin units, with each unit having its own control room. The TID-4844 source terms and associated methodology are used. The USNRC RADTRAD v3.03 code is employed for the transport calculation of radioactive materials in different paths, including control room in-leakage path. The radiological criterion on protection of the operators specified in 10 CFR 50, Appendix A, General Design Criterion 19 is followed. It's demonstrated that the performance of Lungmen control room with 500 cfm unfiltered in-leakage air could meet the radiological habitability acceptance criteria in case of radiation hazards. (author)

Global physical and numerical stability of a nuclear reactor core

International Nuclear Information System (INIS)

Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

2005-01-01

Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

International solar-terrestrial physics program: a plan for the core spaceflight missions

Energy Technology Data Exchange (ETDEWEB)

1985-01-01

This brochure has been prepared to describe the scope of the science problems to be investigated and the mission plan for the core International Solar-Terrestrial Physics (ISTP) Program. This information is intended to stimulate discussions and plans for the comprehensive worldwide ISTP Program. The plan for the study of the solar - terrestrial system is included. The Sun, geospace, and Sun-Earth interaction is discussed as is solar dynamics and the origins of solar winds.

Core Values | NREL

Science.gov (United States)

Core Values Core Values NREL's core values are rooted in a safe and supportive work environment guide our everyday actions and efforts: Safe and supportive work environment Respect for the rights physical and social environment Integrity Maintain the highest standard of ethics, honesty, and integrity

Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

International Nuclear Information System (INIS)

Ikehara, T.; Tsuiki, M.; Takeshita, T.

1990-01-01

On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

International Nuclear Information System (INIS)

Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

2010-01-01

According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

Effects of core strength training using stable versus unstable surfaces on physical fitness in adolescents: a randomized controlled trial.

Science.gov (United States)

Granacher, Urs; Schellbach, Jörg; Klein, Katja; Prieske, Olaf; Baeyens, Jean-Pierre; Muehlbauer, Thomas

2014-01-01

It has been demonstrated that core strength training is an effective means to enhance trunk muscle strength (TMS) and proxies of physical fitness in youth. Of note, cross-sectional studies revealed that the inclusion of unstable elements in core strengthening exercises produced increases in trunk muscle activity and thus provide potential extra training stimuli for performance enhancement. Thus, utilizing unstable surfaces during core strength training may even produce larger performance gains. However, the effects of core strength training using unstable surfaces are unresolved in youth. This randomized controlled study specifically investigated the effects of core strength training performed on stable surfaces (CSTS) compared to unstable surfaces (CSTU) on physical fitness in school-aged children. Twenty-seven (14 girls, 13 boys) healthy subjects (mean age: 14 ± 1 years, age range: 13-15 years) were randomly assigned to a CSTS (n = 13) or a CSTU (n = 14) group. Both training programs lasted 6 weeks (2 sessions/week) and included frontal, dorsal, and lateral core exercises. During CSTU, these exercises were conducted on unstable surfaces (e.g., TOGU© DYNAIR CUSSIONS, THERA-BAND© STABILITY TRAINER). Significant main effects of Time (pre vs. post) were observed for the TMS tests (8-22%, f = 0.47-0.76), the jumping sideways test (4-5%, f = 1.07), and the Y balance test (2-3%, f = 0.46-0.49). Trends towards significance were found for the standing long jump test (1-3%, f = 0.39) and the stand-and-reach test (0-2%, f = 0.39). We could not detect any significant main effects of Group. Significant Time x Group interactions were detected for the stand-and-reach test in favour of the CSTU group (2%, f = 0.54). Core strength training resulted in significant increases in proxies of physical fitness in adolescents. However, CSTU as compared to CSTS had only limited additional effects (i.e., stand-and-reach test). Consequently, if the

The American Society for Radiation Oncology's 2015 Core Physics Curriculum for Radiation Oncology Residents

Energy Technology Data Exchange (ETDEWEB)

Burmeister, Jay, E-mail: burmeist@karmanos.org [Department of Oncology, Karmanos Cancer Center/Wayne State University, Detroit, Michigan (United States); Chen, Zhe [Department of Therapeutic Radiology, Yale University, New Haven, Connecticut (United States); Chetty, Indrin J. [Department of Radiation Oncology, Henry Ford Hospital, Detroit, Michigan (United States); Dieterich, Sonja [Department of Radiation Oncology, University of California – Davis, Sacramento, California (United States); Doemer, Anthony [Department of Radiation Oncology, Henry Ford Hospital, Detroit, Michigan (United States); Dominello, Michael M. [Department of Oncology, Karmanos Cancer Center/Wayne State University, Detroit, Michigan (United States); Howell, Rebecca M. [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas (United States); McDermott, Patrick [Department of Radiation Oncology, Beaumont Health, Royal Oak, Michigan (United States); Nalichowski, Adrian [Karmanos Cancer Center, Detroit, Michigan (United States); Prisciandaro, Joann [Department of Radiation Oncology, University of Michigan, Ann Arbor, Michigan (United States); Ritter, Tim [VA Ann Arbor Healthcare and the University of Michigan, Ann Arbor, Michigan (United States); Smith, Chadd [Department of Radiation Oncology, Henry Ford Hospital, Detroit, Michigan (United States); Schreiber, Eric [Department of Radiation Oncology, University of North Carolina, Chapel Hill, North Carolina (United States); Shafman, Timothy [21st Century Oncology, Fort Myers, Florida (United States); Sutlief, Steven [Department of Radiation Oncology, University of California – San Diego, La Jolla, California (United States); Xiao, Ying [Department of Radiation Oncology, Thomas Jefferson University, Philadelphia, Pennsylvania (United States)

2016-07-15

Purpose: The American Society for Radiation Oncology (ASTRO) Physics Core Curriculum Subcommittee (PCCSC) has updated the recommended physics curriculum for radiation oncology resident education to improve consistency in teaching, intensity, and subject matter. Methods and Materials: The ASTRO PCCSC is composed of physicists and physicians involved in radiation oncology residency education. The PCCSC updated existing sections within the curriculum, created new sections, and attempted to provide additional clinical context to the curricular material through creation of practical clinical experiences. Finally, we reviewed the American Board of Radiology (ABR) blueprint of examination topics for correlation with this curriculum. Results: The new curriculum represents 56 hours of resident physics didactic education, including a 4-hour initial orientation. The committee recommends completion of this curriculum at least twice to assure both timely presentation of material and re-emphasis after clinical experience. In addition, practical clinical physics and treatment planning modules were created as a supplement to the didactic training. Major changes to the curriculum include addition of Fundamental Physics, Stereotactic Radiosurgery/Stereotactic Body Radiation Therapy, and Safety and Incidents sections, and elimination of the Radiopharmaceutical Physics and Dosimetry and Hyperthermia sections. Simulation and Treatment Verification and optional Research and Development in Radiation Oncology sections were also added. A feedback loop was established with the ABR to help assure that the physics component of the ABR radiation oncology initial certification examination remains consistent with this curriculum. Conclusions: The ASTRO physics core curriculum for radiation oncology residents has been updated in an effort to identify the most important physics topics for preparing residents for careers in radiation oncology, to reflect changes in technology and practice since

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

Energy Technology Data Exchange (ETDEWEB)

Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2007-12-15

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

The evaluation of equipment and Instrumentation Reliability Factors on Power Reactor

International Nuclear Information System (INIS)

Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Santosa, Kussigit; Darlis; Sudiyono, Bambang; Yuniyanta, Sasongko; Sudarmin

1999-01-01

Equipment and instrumentation reliability on type power reactor control room was determined by its pattern and design. the principle of ergonomy applied on equipment and instrumentation layout in this ABWR type reactor are geometric pattern appropriate with economic body motion, average anthropometry data of operator especially operator hand-reach, range of vision, angle of vision, lighting, color arrangement and harmony as will as operator case in operating the equipment system. Limitation criteria of the parameter mentioned above are based on EPRI NP-3659, NURG 0700, and NUREG/CR-3331 documents. Besides that, the (working) physical environment parameter factor of the control room must be designed in order to fulfil the standard criteria of ergonomic condition based on NUREG-0800. The reliability evaluation of equipment and instrumentation system also occurs observed from man machine interaction side which happen between operator and equipment and instrumentation in the ABWR type power reactor control room. From the MMI analysis can be known the working failure possibility which is caused by the operator. The evaluation result of equipment and instrumentation reliability on ABWR type power reactor control room showed that the design of this ABWR control room is good and fulfils the ergonomy standard criteria have been determined

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

Physical property control in core/shell inorganic nanostructures for fluorescence and magnetic targeting applications

Science.gov (United States)

Roberts, Stephen K.

Nanomaterials show immense promise for the future in numerous areas of application. Properties that are unique from the bulk material and are tunable allow for innovation in material design. This thesis will focus on controlling the physical properties of core/shell nanostructures to enhance the utility of the materials. The first focus is on the impact of different solvent mixtures during the shell growth phase of SILAR based core/shell quantum dot synthesis is studied. Gaining insight into the mechanism for SILAR growth of core/shell nanoparticles allows improved synthetic yields and precursor binding, providing enhanced control to synthesis of core/shell nanoparticles. The second focus of this thesis is exploring the use of magnetic nanoparticles for magnetic drug targeting for cardiovascular conditions. Magnetic targeting for drug delivery enables increased local drug concentration, while minimizing non-specific interactions. In order to be effective for magnetic targeting, it must be shown that low magnetic strength is sufficient to capture flowing nanoparticles. By demonstrating the binding of a therapeutic agent to the surface at medicinal levels, the viability for use as a nanoparticle drug delivery system is improved.

Conceptual research on reactor core physics for accelerator driven sub-critical reactor

International Nuclear Information System (INIS)

Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

2000-01-01

The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

Does it pay to be smart, attractive, or confident (or all three)? Relationships among general mental ability, physical attractiveness, core self-evaluations, and income.

Science.gov (United States)

Judge, Timothy A; Hurst, Charlice; Simon, Lauren S

2009-05-01

The authors investigated core self-evaluations and educational attainment as mediating mechanisms for the influence of appearance (physical attractiveness) and intelligence (general mental ability) on income and financial strain. The direct effects of core self-evaluations on financial strain, as well as the indirect effects through income, were also considered. Longitudinal data were obtained as part of a national study, the Harvard Study of Health and Life Quality, and proposed models were evaluated with structural equation modeling. Results supported a partially mediated model, such that general mental ability and physical attractiveness exhibited both direct and indirect effects on income, as mediated by educational attainment and core self-evaluations. Finally, income negatively predicted financial strain, whereas core self-evaluations had both a direct and an indirect (through income) negative effect on financial strain. Overall, the results suggest that looks (physical attractiveness), brains (intelligence), and personality (core self-evaluations) are all important to income and financial strain. (c) 2009 APA, all rights reserved.

The ultimate response guideline simulation and analysis by using (TRACE) for Lungmen ABWR nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lin, Hao-Tzu [Atomic Energy Council, Inst. of Nuclear Energy Research, Taoyuan City, Taiwan (China); Yang, Shu-Ming; Chen, Shao-Wen; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Inst. of Nuclear Engineering and Science, HsinChu, Taiwan (China)

2015-07-15

In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

Physical and chemical characteristics of L1689-SMM16, an oscillating prestellar core in Ophiuchus

International Nuclear Information System (INIS)

Chitsazzadeh, S.; Di Francesco, J.; Sadavoy, S. I.; Schnee, S.; Friesen, R. K.; Shimajiri, Y.; Langston, G. I.; Bourke, T. L.; Keto, E. R.; Pineda, J. E.; Takakuwa, S.; Tatematsu, K.

2014-01-01

We present single-dish observations of the L1689-SMM16 core in the Ophiuchus molecular cloud in NH 3 (1, 1) and (2, 2) emission using the Green Bank Telescope, in N 2 H + (1-0) emission using the Nobeyama Radio Observatory, and in NH 2 D (1 1,1 a (--)1 0,1 s ), HCN (1-0), HNC (1-0), H 13 CO + (1-0), and HCO + (1-0) emission using the Mopra telescope. The morphologies of the integrated NH 3 (1, 1) and N 2 H + (1-0) emission well match that of 250 μm continuum emission. Line widths of NH 3 (1, 1) and N 2 H + (1-0) show the presence of transonic turbulence across the core. Jeans and virial analyses made using updated measurements of core mass and size confirm that L1689-SMM16 is prestellar, i.e., gravitationally bound. It also has accumulated more mass compared to its corresponding Jeans mass in the absence of magnetic fields and therefore is a 'super-Jeans' core. The high levels of X(NH 3 )/X(N 2 H + ) and deuterium fractionation reinforce the idea that the core has not yet formed a protostar. Comparing the physical parameters of the core with those of a Bonnor-Ebert sphere reveals the advanced evolutionary stage of L1689-SMM16 and shows that it might be unstable to collapse. We do not detect any evidence of infall motions toward the core. Instead, red asymmetry in the line profiles of HCN (1-0) and HNC (1-0) indicates the expansion of the outer layers of the core at a speed of ∼0.2 km s –1 to 0.3 km s –1 . For a gravitationally bound core, expansion in the outer layers might indicate that the core is experiencing oscillations.

ABWR3. Passive double confinement and SA population dose evaluation for the innovative ABWR

International Nuclear Information System (INIS)

Matsumoto, Keiji; Hosomi, Kenji; Sato, Takashi

2015-01-01

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. The iB1350 uses the Mark W containment and the in-containment filtered venting system (IFVS). The Mark W containment is made of reinforced concrete and has double cylinder FP barriers. There are also IC/PCCS pools, a fuel pool and a dryer separator (DS) pool on the top slab of the containment. These pools work as water seal for FP leakage through the top slab. Most FP such as CsI are scrubbed in the pools. The containment head is also submerged in the reactor well pool. The pool water works as shielding for radiation from the reactor core during normal operation and water seal for FP scrubbing dining a severe accident. The base mat concrete is covered with the S/P and the core catcher that is also submerged with corium flooding water during an accident. Therefore, the Mark W containment has passive double confinement barriers for FP. Moreover, the IFVS works as in-containment filtered venting system that scrubs FP from the wet well (WW) and the dry well (DW). The filtered venting tank is arranged in the outer well (OW) of the Mark W containment. Even noble gases and organic iodine going through the filtered venting system are still confined inside the containment and never released directly to the environment. The IFVS uses the innovative passive containment cooling system (iPCCS) as the pre-stage heat removal system. The iPCCS has a normal open suction line from the WW. Therefore, FP are scrubbed in the S/P at first and then vented into the filtered venting tank. After the DW suction line of the iPCCS is opened all the steam is cooled and condensed in the heat exchanger. Most FP are trapped in the condensate and returned into the WW through the condensate return line of the iPCCS. Therefore, the Mark W containment has excellent FP double confinement barriers and the in-containment filtered venting system that enable

Engineered inorganic core/shell nanoparticles

Energy Technology Data Exchange (ETDEWEB)

Mélinon, Patrice, E-mail: patrice.melinon@univ-lyon1.fr [Institut Lumière matière Université Claude Bernard Lyon 1 et CNRS et OMNT, Domaine Scientifique de la Doua, Bâtiment Léon Brillouin, 43 Boulevard du 11 Novembre 1918, F 69622 Villeurbanne (France); Begin-Colin, Sylvie [IPCMS et OMNT, 23 rue du Loess BP 43, 67034 STRASBOURG Cedex 2 (France); Duvail, Jean Luc [IMN UMR 6502 et OMNT Campus Sciences : 2 rue de la Houssinire, BP32229, 44322 Nantes Cedex3 (France); Gauffre, Fabienne [SPM et OMNT : Institut des sciences chimiques de Rennes - UMR 6226, 263 Avenue du General Leclerc, CS 74205, 35042 RENNES Cedex (France); Boime, Nathalie Herlin [IRAMIS-NIMBE, Laboratoire Francis Perrin (CEA CNRS URA 2453) et OMNT, Bat 522, CEA Saclay, 91191 Gif sur Yvette Cedex (France); Ledoux, Gilles [Institut Lumière Matière Université Claude Bernard Lyon 1 et CNRS et OMNT, Domaine Scientifique de la Doua, Bâtiment Alfred Kastler 43 Boulevard du 11 Novembre 1918 F 69622 Villeurbanne (France); Plain, Jérôme [Universit de technologie de Troyes LNIO-ICD, CNRS et OMNT 12 rue Marie Curie - CS 42060 - 10004 Troyes cedex (France); Reiss, Peter [CEA Grenoble, INAC-SPrAM, UMR 5819 CEA-CNRS-UJF et OMNT, Grenoble cedex 9 (France); Silly, Fabien [CEA, IRAMIS, SPEC, TITANS, CNRS 2464 et OMNT, F-91191 Gif sur Yvette (France); Warot-Fonrose, Bénédicte [CEMES-CNRS, Université de Toulouse et OMNT, 29 rue Jeanne Marvig F 31055 Toulouse (France)

2014-10-20

It has been for a long time recognized that nanoparticles are of great scientific interest as they are effectively a bridge between bulk materials and atomic structures. At first, size effects occurring in single elements have been studied. More recently, progress in chemical and physical synthesis routes permitted the preparation of more complex structures. Such structures take advantages of new adjustable parameters including stoichiometry, chemical ordering, shape and segregation opening new fields with tailored materials for biology, mechanics, optics magnetism, chemistry catalysis, solar cells and microelectronics. Among them, core/shell structures are a particular class of nanoparticles made with an inorganic core and one or several inorganic shell layer(s). In earlier work, the shell was merely used as a protective coating for the core. More recently, it has been shown that it is possible to tune the physical properties in a larger range than that of each material taken separately. The goal of the present review is to discuss the basic properties of the different types of core/shell nanoparticles including a large variety of heterostructures. We restrict ourselves on all inorganic (on inorganic/inorganic) core/shell structures. In the light of recent developments, the applications of inorganic core/shell particles are found in many fields including biology, chemistry, physics and engineering. In addition to a representative overview of the properties, general concepts based on solid state physics are considered for material selection and for identifying criteria linking the core/shell structure and its resulting properties. Chemical and physical routes for the synthesis and specific methods for the study of core/shell nanoparticle are briefly discussed.

GENOVA: a generalized perturbation theory program for various applications to CANDU core physics analysis (I)-theory and application

International Nuclear Information System (INIS)

Kim, Do Heon; Choi, Hang Bok

2001-01-01

A generalized perturbation theory (GPT) program, GENOVA, has been developed for the purpose of various applications to Canadian deuterium uranium (CANDU) reactor physics analyses. GENOVA was written under the framework of CANDU physics design and analysis code, RFSP. A sensitivity method based on the GPT was implemented in GENOVA to estimate various sensitivity coefficients related to the movement of zone controller units (ZCUs) existing in the CANDU reactor. The numerical algorithm for the sensitivity method was verified by a simple 2 x 2 node problem. The capability of predicting ZCU levels upon a refueling perturbation was validated for a CANDU-6 reactor problem. The applicability of GENOVA to the CANDU-6 core physics analysis has been demonstrated with the optimum refueling simulation and the uncertainty analysis problems. For the optimum refueling simulation, an optimum channel selection strategy has been proposed, using the ZCU level predicted by GENOVA. The refueling simulation of a CANDU-6 natural uranium core has shown that the ZCU levels are successfully controlled within the operating range while the channel and bundle powers are satisfying the license limits. An uncertainty analysis has been performed for the fuel composition heterogeneity of a CANDU DUPIC core, using the sensitivity coefficients generated by GENOVA. The results have shown that the uncertainty of the core performance parameter can be reduced appreciably when the contents of the major fissile isotopes are tightly controlled. GENOVA code has been successfully explored to supplement the weak points of the current design and analysis code, such as the incapacity of performing an optimum refueling simulation and uncertainty analysis. The sample calculations have shown that GENOVA has strong potential to be used for CANDU core analysis combined with the current design and analysis code, RFSP, especially for the development of advanced CANDU fuels

Analysis of Core Stability Exercise Effect on the Physical and Psychological Function of Elderly Women Vulnerable to Falls during Obstacle Negotiation.

Science.gov (United States)

Ko, Dae-Sik; Jung, Dae-In; Jeong, Mi-Ae

2014-11-01

[Purpose] The aim of the present study was to investigate the effects of core stability exercise (CSE) on the physical and psychological functions of elderly women while negotiating general obstacles. [Subjects and Methods] After allocating 10 elderly women each to the core stability training group and the control group, we carried out Performance-Oriented Mobility Assessment (POMA) and measured crossing velocity (CV), maximum vertical heel clearance (MVHC), and knee flexion angle for assessing physical performances. We evaluated depression and fear of falling for assessing psychological functions. [Results] Relative to the control group, the core stability training group showed statistically significant overall changes after the training session: an increase in POMA scores, faster CV, lower MVHC, and a decrease in knee flexion angle. Furthermore, depression and fear of falling decreased significantly. [Conclusion] CSE can have a positive effect on the improvement of physical and psychological performances of older women who are vulnerable to falls as they negotiate everyday obstacles.

Evaluation of core physics analysis methods for conversion of the INL advanced test reactor to low-enrichment fuel

International Nuclear Information System (INIS)

DeHart, M. D.; Chang, G. S.

2012-01-01

Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR. (authors)

Physical Layer Multi-Core Prototyping A Dataflow-Based Approach for LTE eNodeB

CERN Document Server

Pelcat, Maxime; Piat, Jonathan; Nezan, Jean-François

2013-01-01

Base stations developed according to the 3GPP Long Term Evolution (LTE) standard require unprecedented processing power. 3GPP LTE enables data rates beyond hundreds of Mbits/s by using advanced technologies, necessitating a highly complex LTE physical layer. The operating power of base stations is a significant cost for operators, and is currently optimized using state-of-the-art hardware solutions, such as heterogeneous distributed systems. The traditional system design method of porting algorithms to heterogeneous distributed systems based on test-and-refine methods is a manual, thus time-expensive, task.   Physical Layer Multi-Core Prototyping: A Dataflow-Based Approach for LTE eNodeB provides a clear introduction to the 3GPP LTE physical layer and to dataflow-based prototyping and programming. The difficulties in the process of 3GPP LTE physical layer porting are outlined, with particular focus on automatic partitioning and scheduling, load balancing and computation latency reduction, specifically in sys...

The Physical Mechanism of Core-Wide and Local Instabilities at the Forsmark-1 BWR

International Nuclear Information System (INIS)

Analytis, G. Th.

1998-10-01

During the last 15 years, the problem of BWR instabilities has attracted the attention of a number of researchers. From the theoretical point of view, one would be interested in physically understanding the mechanisms responsible for the in- and out-of-phase core wide power oscillations observed at certain operating points of the power-flow map in different BWRs. From the practical point of view, one must try to avoid these 'incidents' since either locally, or globally, the power may substantially exceed the prescribed levels. In this work, we shall use RAMONA3-12 and analyse a rather unusual instability incident at Forsmark-1 in which in addition to the core-wide fundamental spatial mode oscillation, there were local large amplitude power oscillations at different radial positions in the core. We were able to reproduce these unusual experimental findings by assuming that there are large amplitude Density Wave Oscillations (DWOs) in different bundles, induced by the fact that these bundles were not seated properly into the lower fuel support plate. (author)

Multi-physics and multi-objective design of heterogeneous SFR core: development of an optimization method under uncertainty

International Nuclear Information System (INIS)

Ammar, Karim

2014-01-01

Since Phenix shutting down in 2010, CEA does not have Sodium Fast Reactor (SFR) in operating condition. According to global energetic challenge and fast reactor abilities, CEA launched a program of industrial demonstrator called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a reactor with electric power capacity equal to 600 MW. Objective of the prototype is, in first to be a response to environmental constraints, in second demonstrates the industrial viability of SFR reactor. The goal is to have a safety level at least equal to 3. generation reactors. ASTRID design integrates Fukushima feedback; Waste reprocessing (with minor actinide transmutation) and it linked industry. Installation safety is the priority. In all cases, no radionuclide should be released into environment. To achieve this objective, it is imperative to predict the impact of uncertainty sources on reactor behaviour. In this context, this thesis aims to develop new optimization methods for SFR cores. The goal is to improve the robustness and reliability of reactors in response to existing uncertainties. We will use ASTRID core as reference to estimate interest of new methods and tools developed. The impact of multi-Physics uncertainties in the calculation of the core performance and the use of optimization methods introduce new problems: How to optimize 'complex' cores (i.e. associated with design spaces of high dimensions with more than 20 variable parameters), taking into account the uncertainties? What is uncertainties behaviour for optimization core compare to reference core? Taking into account uncertainties, optimization core are they still competitive? Optimizations improvements are higher than uncertainty margins? The thesis helps to develop and implement methods necessary to take into account uncertainties in the new generation of simulation tools. Statistical methods to ensure consistency of complex multi-Physics simulation results are also

Physical and chemical characteristics of L1689-SMM16, an oscillating prestellar core in Ophiuchus

Energy Technology Data Exchange (ETDEWEB)

Chitsazzadeh, S.; Di Francesco, J.; Sadavoy, S. I. [Department of Physics and Astronomy, The University of Victoria, Victoria, BC V8P 5C2 (Canada); Schnee, S. [National Radio Astronomy Observatory, 520 Edgemont Road, Charlottesville, VA 22903 (United States); Friesen, R. K. [The Dunlap Institute for Astronomy and Astrophysics, University of Toronto, 50 St. George St., Toronto, ON M5S 3H4 (Canada); Shimajiri, Y. [Laboratoire AIM, CEA/DSM-CNRS-Université Paris Diderot, IRFU/Service d' Astrophysique, CEA Saclay, F-91191 Gif-sur-Yvette (France); Langston, G. I. [National Radio Astronomy Observatory, P.O. Box 2, Green Bank, WV 24944 (United States); Bourke, T. L.; Keto, E. R. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Pineda, J. E. [Institute for Astronomy, ETH Zurich, Wolfgang-Pauli-Strasse 27, CH-8093 Zurich (Switzerland); Takakuwa, S. [Academia Sinica Institute of Astronomy and Astrophysics, P.O. Box 23-141, Taipei 10617, Taiwan (China); Tatematsu, K., E-mail: schitsaz@uvic.ca [National Astronomical Observatory of Japan, 2-21-1 Osawa, Mitaka, Tokyo 181-8588 (Japan)

2014-08-01

We present single-dish observations of the L1689-SMM16 core in the Ophiuchus molecular cloud in NH{sub 3} (1, 1) and (2, 2) emission using the Green Bank Telescope, in N{sub 2}H{sup +} (1-0) emission using the Nobeyama Radio Observatory, and in NH{sub 2}D (1{sub 1,1}{sup a}(--)1{sub 0,1}{sup s}), HCN (1-0), HNC (1-0), H{sup 13}CO{sup +} (1-0), and HCO{sup +} (1-0) emission using the Mopra telescope. The morphologies of the integrated NH{sub 3} (1, 1) and N{sub 2}H{sup +} (1-0) emission well match that of 250 μm continuum emission. Line widths of NH{sub 3} (1, 1) and N{sub 2}H{sup +} (1-0) show the presence of transonic turbulence across the core. Jeans and virial analyses made using updated measurements of core mass and size confirm that L1689-SMM16 is prestellar, i.e., gravitationally bound. It also has accumulated more mass compared to its corresponding Jeans mass in the absence of magnetic fields and therefore is a 'super-Jeans' core. The high levels of X(NH{sub 3})/X(N{sub 2}H{sup +}) and deuterium fractionation reinforce the idea that the core has not yet formed a protostar. Comparing the physical parameters of the core with those of a Bonnor-Ebert sphere reveals the advanced evolutionary stage of L1689-SMM16 and shows that it might be unstable to collapse. We do not detect any evidence of infall motions toward the core. Instead, red asymmetry in the line profiles of HCN (1-0) and HNC (1-0) indicates the expansion of the outer layers of the core at a speed of ∼0.2 km s{sup –1} to 0.3 km s{sup –1}. For a gravitationally bound core, expansion in the outer layers might indicate that the core is experiencing oscillations.

Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

International Nuclear Information System (INIS)

Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

1994-01-01

The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs

Nuclear physics: the core of matter, the fuel of stars

International Nuclear Information System (INIS)

Schiffer, J.P.

1999-01-01

Dramatic progress has been made in all branches of physics since the National Research Council's 1986 decadal survey of the field. The Physics in a New Era series explores these advances and looks ahead to future goals. The series includes assessments of the major subfields and reports on several smaller subfields, and preparation has begun on an overview volume on the unity of physics, its relationships to other fields, and its contributions to national needs. Nuclear Physics is the latest volume of the series. The book describes current activity in understanding nuclear structure and symmetries, the behavior of matter at extreme densities, the role of nuclear physics in astrophysics and cosmology, and the instrumentation and facilities used by the field. It makes recommendations on the resources needed for experimental and theoretical advances in the coming decade. Nuclear physics addresses the nature of matter making up 99.9 percent of the mass of our everyday world. It explores the nuclear reactions that fuel the stars, including our Sun, which provides the energy for all life on Earth. The field of nuclear physics encompasses some 3,000 experimental and theoretical researchers who work at universities and national laboratories across the United States, as well as the experimental facilities and infrastructure that allow these researchers to address the outstanding scientific questions facing us. This report provides an overview of the frontiers of nuclear physics as we enter the next millennium, with special attention to the state of the science in the United States.The current frontiers of nuclear physics involve fundamental and rapidly evolving issues. One is understanding the structure and behavior of strongly interacting matter in terms of its basic constituents, quarks and gluons, over a wide range of conditions - from normal nuclear matter to the dense cores of neutron stars, and to the Big Bang that was the birth of the universe. Another is to describe

ASTRO's 2007 Core Physics Curriculum for Radiation Oncology Residents

International Nuclear Information System (INIS)

Klein, Eric E.; Gerbi, Bruce J.; Price, Robert A.; Balter, James M.; Paliwal, Bhudatt; Hughes, Lesley; Huang, Eugene

2007-01-01

In 2004, American Society for Therapeutic Radiology and Oncology (ASTRO) published a curriculum for physics education. The document described a 54-hour course. In 2006, the committee reconvened to update the curriculum. The committee is composed of physicists and physicians from various residency program teaching institutions. Simultaneously, members have associations with American Association of Physicists in Medicine, ASTRO, Association of Residents in Radiation Oncology, American Board of Radiology, and American College of Radiology. Representatives from the latter two organizations are key to provide feedback between the examining organizations and ASTRO. Subjects are based on Accreditation Council for Graduate Medical Education requirements (particles and hyperthermia), whereas the majority of subjects and appropriated hours/subject were developed by consensus. The new curriculum is 55 hours, containing new subjects, redistribution of subjects with updates, and reorganization of core topics. For each subject, learning objectives are provided, and for each lecture hour, a detailed outline of material to be covered is provided. Some changes include a decrease in basic radiologic physics, addition of informatics as a subject, increase in intensity-modulated radiotherapy, and migration of some brachytherapy hours to radiopharmaceuticals. The new curriculum was approved by the ASTRO board in late 2006. It is hoped that physicists will adopt the curriculum for structuring their didactic teaching program, and simultaneously, American Board of Radiology, for its written examination. American College of Radiology uses the ASTRO curriculum for their training examination topics. In addition to the curriculum, the committee added suggested references, a glossary, and a condensed version of lectures for a Postgraduate Year 2 resident physics orientation. To ensure continued commitment to a current and relevant curriculum, subject matter will be updated again in 2 years

Reactor core fuel management

International Nuclear Information System (INIS)

Silvennoinen, P.

1976-01-01

The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

International Nuclear Information System (INIS)

1997-05-01

This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff's review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff's review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design

Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-05-01

This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7

Directory of Open Access Journals (Sweden)

Daniel J. Kelly, III

2017-09-01

Full Text Available The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of 0.99994 ± 6.8E-6 (95% confidence interval. Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

SF36 physical functioning scale and 2-minute walk test advocated as core qualifiers to evaluate physical functioning in patients with late-onset sequelae of poliomyelitis

NARCIS (Netherlands)

Stolwijk-Swüste, Janneke M.; Beelen, Anita; Lankhorst, Gustaaf J.; Nollet, Frans; Stolwijk-Swüste, J. M.; Lankhorst, G. J.; Dekker, J.; van Dijk, G. M.; van den Ende, C. H. M.; Post, B.; de Haan, R. J.; Speelman, H.

2008-01-01

OBJECTIVE: To select a questionnaire and walking capacity test based on comparison of clinimetric properties and mutual association to be used as "core" qualifiers for physical functioning in patients with late-onset sequelae of poliomyelitis. DESIGN: Repeated-measures at 3-week intervals. SUBJECTS:

Sf36 physical functioning scale and 2-minute walk test advocated as core qualifiers to evaluate physical functioning in patients with late-onset sequelae of poliomyelitis.

NARCIS (Netherlands)

Stolwijk-Swuste, J.M.; Beelen, J.A.J.M.; Lankhorst, G.J.; Nollet, F.

2008-01-01

Objective: To select a questionnaire and walking capacity test based on comparison of clinimetric properties and mutual association to be used as "core" qualifiers for physical functioning in patients with late-onset sequelae of poliomyelitis. Design: Repeated-measures at 3-week intervals. Subjects:

Methodology for reactor core physics analysis - part 2; Metodologia de analise fisica do nucleo - etapa 2

Energy Technology Data Exchange (ETDEWEB)

Ponzoni Filho, P; Fernandes, V B; Lima Bezerra, J de; Santos, T I.C.

1992-12-01

The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs.

A core physical examination in internal medicine: what should students do and how about their supervisors?

Science.gov (United States)

Haring, Catharina M; van der Meer, Jos W M; Postma, Cornelis T

2013-09-01

Performance of a focused physical examination will induce a high cognitive load for medical students in the early phase of the clinical clerkships. To come to a workable and clinically applicable standard physical examination for medical students to be used in every new patient in the daily clinical practice of internal medicine. A questionnaire held among physicians that supervise students during the clerkship of internal medicine in one Dutch training region. Of the complete list of physical examination 55 items were considered to be an integral part of the standard general physical examination for medical students. Most emphasized were elements of the physical examination aimed at general parameters, thorax and abdomen, vascular status, lymph nodes, spinal column, skin and some parts of the neurological examination. The standard physical examinations performed by supervisors themselves contain fewer items than they expected from the students. The expectations a supervisor has towards the student correlates with the frequency with which they apply the various components in their own physical examination. This study provides us with a 'core' physical examination for medical students that can be applied in the early phase of the clinical clerkships.

Design Report for the core design of the first core Mark-Ia of the SNR-300

International Nuclear Information System (INIS)

Stanculescu, A.

1984-05-01

The report describes the first core Mark-Ia of the SNR-300 reactor and its different assembly types with their operational strategy. Methods, criteria and results of the neutron physical, thermal hydraulic and core mechanical design of the whole core and its assemblies are presented

Maximum stellar iron core mass

Indian Academy of Sciences (India)

60, No. 3. — journal of. March 2003 physics pp. 415–422. Maximum stellar iron core mass. F W GIACOBBE. Chicago Research Center/American Air Liquide ... iron core compression due to the weight of non-ferrous matter overlying the iron cores within large .... thermal equilibrium velocities will tend to be non-relativistic.

76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

Science.gov (United States)

2011-01-20

... accepted the application as a docketed application for amendment to the U.S. ABWR design certification... perform the design work associated with the amended portion of the U.S. ABWR design represented by STPNOC's application and to supply the amended portion of the U.S. ABWR design. STPNOC's amendment to the U...

Impact of advanced BWR core physics method on BWR core monitoring

Energy Technology Data Exchange (ETDEWEB)

Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

2000-07-01

Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

Core design methods for advanced LMFBRs

International Nuclear Information System (INIS)

Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

1977-05-01

The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

Energy Technology Data Exchange (ETDEWEB)

Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kondo, Toshiki [Graduate School of Engineering, Osaka University (Japan); Ishikawa, Takehiko [Japan Aerospace Exploration Agency (Japan); SOKEN-DAI (Graduate University for Advanced Studies) (Japan); Okada, Junpei T. [Institute for Materials Research, Tohoku University (Japan); Watanabe, Yuki [Advanced Engineering Services Co. Ltd. (Japan); Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

2017-03-15

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 0.77}Cr{sub 0.23}) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 77}Cr{sub 23}. • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

International Nuclear Information System (INIS)

Ohishi, Yuji; Kondo, Toshiki; Ishikawa, Takehiko; Okada, Junpei T.; Watanabe, Yuki; Muta, Hiroaki; Kurosaki, Ken; Yamanaka, Shinsuke

2017-01-01

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 0.77 Cr 0.23 ) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 77 Cr 23 . • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

Using robustness and preferred locations of archeomagnetic flux patches to constrain the physics of the core

Science.gov (United States)

Terra-Nova, F.; Amit, H.; Hartmann, G. A.; Trindade, R. I. F.

2017-12-01

Archaeomagnetic field models cover longer timescales than historical models and may therefore resolve the motion of geomagnetic features on the core-mantle boundary (CMB) in a more meaningful statistical sense. Here we perform a detailed appraisal of archaeomagnetic field models to infer some aspects of the physics of the outer core. We characterize and compare the identification and tracking of reversed flux patches (RFPs) in order to assess the RFPs robustness. We find similar behaviour within a family of models but differences among different families, suggesting that modelling strategy is more influential than data set. Similarities involve recurrent positions of RFPs, but no preferred direction of motion is found. The tracking of normal flux patches shows similar qualitative behaviour confirming that RFPs identification and tracking is not strongly biased by their relative weakness. We also compare the tracking of RFPs with that of the historical field model gufm1 and with seismic anomalies of the lowermost mantle to explore the possibility that RFPs have preferred locations prescribed by lower mantle lateral heterogeneity. The archaeomagnetic field model that most resembles the historical field is interpreted in terms of core dynamics and core-mantle thermal interactions. This model exhibits correlation between RFPs and low seismic shear velocity in co-latitude and a shift in longitude. These results shed light on core processes, in particular we infer toroidal field lines with azimuthal orientation below the CMB and large fluid upwelling structures with a width of about 80° (Africa) and 110° (Pacific) at the top of the core. Finally, similar preferred locations of RFPs in the past 9 and 3 kyr of the same archaeomagnetic field model suggest that a 3 kyr period is sufficiently long to reliably detect mantle control on core dynamics. This allows estimating an upper bound of 220-310 km for the magnetic boundary layer thickness below the CMB.

Sediment Core Laboratory

Data.gov (United States)

Federal Laboratory Consortium — FUNCTION: Provides instrumentation and expertise for physical and geoacoustic characterization of marine sediments.DESCRIPTION: The multisensor core logger measures...

Development and Integration of Professional Core Values Among Practicing Clinicians.

Science.gov (United States)

McGinnis, Patricia Quinn; Guenther, Lee Ann; Wainwright, Susan F

2016-09-01

The physical therapy profession has adopted professional core values, which define expected values for its members, and developed a self-assessment tool with sample behaviors for each of the 7 core values. However, evidence related to the integration of these core values into practice is limited. The aims of this study were: (1) to gain insight into physical therapists' development of professional core values and (2) to gain insight into participants' integration of professional core values into clinical practice. A qualitative design permitted in-depth exploration of the development and integration of the American Physical Therapy Association's professional core values into physical therapist practice. Twenty practicing physical therapists were purposefully selected to explore the role of varied professional, postprofessional, and continuing education experiences related to exposure to professional values. The Core Values Self-Assessment and résumé sort served as prompts for reflection via semistructured interviews. Three themes were identified: (1) personal values were the foundation for developing professional values, which were further shaped by academic and clinical experiences, (2) core values were integrated into practice independent of practice setting and varied career paths, and (3) participants described the following professional core values as well integrated into their practice: integrity, compassion/caring, and accountability. Social responsibility was an area consistently identified as not being integrated into their practice. The Core Values Self-Assessment tool is a consensus-based document developed through a Delphi process. Future studies to establish reliability and construct validity of the tool may be warranted. Gaining an in-depth understanding of how practicing clinicians incorporate professional core values into clinical practice may shed light on the relationship between core values mastery and its impact on patient care. Findings may

Reactor physics

International Nuclear Information System (INIS)

Ait Abderrahim, H.

1998-01-01

Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

Energy Technology Data Exchange (ETDEWEB)

Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

2014-10-15

The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

Activities of working party on 'Subcritical core of accelerator-driven system' under the research committee on reactor physics of AESJ and JAERI

International Nuclear Information System (INIS)

Iwasaki, T.; Tsujimoto, K.; Nishihara, K.; Kitamura, Y.

2004-01-01

The Research Committee on Reactor Physics under the Atomic Energy Society of Japan and the Japan Atomic Energy Research Inst. organized the working party (ADS-WP) on S ubcritical Core of Accelerator-Driven System . The ADS-WP investigated reactor physics of subcriticality from the viewpoint of the accelerator driven system (ADS) since subcriticality has been almost studied from the viewpoint of critical safety. The working party was set in July 2001 and it worked for two years. The activities of the ADS-WP are (Work-I) theory of subcriticality, (Work-II) benchmark of subcritical core, (Work-III) setting of subcriticality level of ADS and (Work-JAO monitoring of subcriticality. These activities clarified about the important issues related to the subcriticality or the subcritical core from the wide ranges of theory, analysis, calculation, design and monitoring for ADS. The activities were already summarized and the report will be published in March 2004. (authors)

Analysis of Core Stability Exercise Effect on the Physical and Psychological Function of Elderly Women Vulnerable to Falls during Obstacle Negotiation

OpenAIRE

Ko, Dae-Sik; Jung, Dae-In; Jeong, Mi-Ae

2014-01-01

[Purpose] The aim of the present study was to investigate the effects of core stability exercise (CSE) on the physical and psychological functions of elderly women while negotiating general obstacles. [Subjects and Methods] After allocating 10 elderly women each to the core stability training group and the control group, we carried out Performance-Oriented Mobility Assessment (POMA) and measured crossing velocity (CV), maximum vertical heel clearance (MVHC), and knee flexion angle for assessi...

TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)

2012-07-01

The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

Soil architecture relationships with dynamic soil physical processes: a conceptual study using natural, artificial, and 3D-printed soil cores

DEFF Research Database (Denmark)

Lamandé, Mathieu; Schjønning, Per; Dal Ferro, Nicola

Pore system architecture is a key feature for understanding physical, biological and chemical processes in soils. Development of visualisation technics, especially x-ray CT, during recent years has been useful in describing the complex relationships between soil architecture and soil functions. We...... believe that combining visualization with physical models is a step further towards a better understanding of these relationships. We conducted a concept study using natural, artificial and 3D-printed soil cores. Eight natural soil cores (100 cm3) were sampled in a cultivated stagnic Luvisol at two depths...... (topsoil and subsoil), representing contrasting soil pore systems. Cylinders (100 cm3) were produced from plastic or from autoclaved aerated concrete. Holes of diameters 1.5 and 3 mm were drilled in the cylinder direction for the plastic cylinder and for one of the AAC cylinders. All natural and artificial...

DENSE CORES IN THE PIPE NEBULA: AN IMPROVED CORE MASS FUNCTION

International Nuclear Information System (INIS)

Rathborne, J. M.; Lada, C. J.; Muench, A. A.; Alves, J. F.; Kainulainen, J.; Lombardi, M.

2009-01-01

In this paper, we derive an improved core mass function (CMF) for the Pipe Nebula from a detailed comparison between measurements of visual extinction and molecular-line emission. We have compiled a refined sample of 201 dense cores toward the Pipe Nebula using a two-dimensional threshold identification algorithm informed by recent simulations of dense core populations. Measurements of radial velocities using complimentary C 18 O (1-0) observations enable us to cull out from this sample those 43 extinction peaks that are either not associated with dense gas or are not physically associated with the Pipe Nebula. Moreover, we use the derived C 18 O central velocities to differentiate between single cores with internal structure and blends of two or more physically distinct cores, superposed along the same line of sight. We then are able to produce a more robust dense core sample for future follow-up studies and a more reliable CMF than was possible previously. We confirm earlier indications that the CMF for the Pipe Nebula departs from a single power-law-like form with a break or knee at M ∼ 2.7 ± 1.3 M sun . Moreover, we also confirm that the CMF exhibits a similar shape to the stellar initial mass function (IMF), but is scaled to higher masses by a factor of ∼4.5. We interpret this difference in scaling to be a measure of the star formation efficiency (22% ± 8%). This supports earlier suggestions that the stellar IMF may originate more or less directly from the CMF.

Preliminary Physical Stratigraphy and Geophysical Data From the USGS Dixon Core, Onslow County, North Carolina

Science.gov (United States)

Seefelt, Ellen L.; Gonzalez, Wilma Aleman B.; Self-Trail, Jean M.; Weems, Robert E.; Edwards, Lucy E.; Pierce, Herbert A.; Durand, Colleen T.

2009-01-01

In October through November 2006, scientists from the U. S. Geological Survey (USGS) Eastern Region Earth Surface Processes Team (EESPT) and the Raleigh (N.C.) Water Science Center (WSC), in cooperation with the North Carolina Geological Survey (NCGS) and the Onslow County Water and Sewer Authority (ONWASA), drilled a stratigraphic test hole and well in Onslow County, N.C. The Dixon corehole was cored on ONWASA water utility property north of the town of Dixon, N.C., in the Sneads Ferry 7.5-minute quadrangle at latitude 34deg33'35' N, longitude 77deg26'54' W (decimal degrees 34.559722 and -77.448333). The site elevation is 66.0 feet (ft) above mean sea level as determined using a Paulin precision altimeter. The corehole attained a total depth of 1,010 ft and was continuously cored by the USGS EESPT drilling crew. A groundwater monitoring well was installed in the screened interval between 234 and 254 ft below land surface. The section cored at this site includes Upper Cretaceous, Paleogene, and Neogene sediments. The Dixon core is stored at the NCGS Coastal Plain core storage facility in Raleigh. The Dixon corehole is the fourth and last in a series of planned North Carolina benchmark coreholes drilled by the USGS Coastal Carolina Project. These coreholes explore the physical stratigraphy, facies, and thickness of Cretaceous, Paleogene, and Neogene Coastal Plain sediments in North Carolina. Correlations of lithologies, facies, and sequence stratigraphy can be made with the Hope Plantation corehole, N.C., near Windsor in Bertie County (Weems and others, 2007); the Elizabethtown corehole, near Elizabethtown, N.C., in Bladen County (Self-Trail and others, 2004b); the Smith Elementary School corehole, near Cove City, N.C., in Craven County (Harris and Self-Trail, 2006; Crocetti, 2007); the Kure Beach corehole, near Wilmington, N.C., in New Hanover County (Self-Trail and others, 2004a); the Esso#1, Esso #2, Mobil #1, and Mobil #2 cores in Albermarle and Pamlico Sounds

World class BWRs for the 1990s

International Nuclear Information System (INIS)

Wilkins, D.; Lowton, R.

1989-01-01

The first step in Advanced BWR development took place in 1978. It was the formation in San Jose, California, of an Advanced Engineering Team, consisting of representatives of worldwide BWR suppliers and leading architect engineer firms. The Advanced Engineering Team effort was completed in 1979 and established the basic ABWR concept. During the period 1980-85, ABWR development was focused in Japan. Under the sponsorship of the Tokyo Electric Power Company and Japanese BWR utilities, GE, Hitachi and Toshiba performed design development studies to confirm the technical and economic attractiveness of the ABWR, and also conducted a comprehensive test programme encompassing the new features. In 1987, The Tokyo Electric Power Company announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa nuclear power station with commercial operation of the first unit in 1996 and the second in 1998. The company submitted the application for the establishment permit for the two ABWR units in May 1988. The units will be supplied by a joint venture of the three companies who worked on the development studies. This article goes on to describe the ABWR design resulting form this international development effort. (3 figures, 1 table). (Author)

Nuclear power for the new millennium

International Nuclear Information System (INIS)

Hucik, S.A.; Redding, J.R.

1998-01-01

Advanced nuclear technology is being commercially deployed. Two ABWR's have been constructed in Japan and are reliably generating large amounts of low cost electricity. Taiwan is now in the process of licensing and constructing two more ABWR's, which will enter commercial operation in 2004 and 2005. Other countries have similar strategies to deploy advanced nuclear plants and the successful deployment of ABWR's in Japan and Taiwan, coupled with international agreements to limit CO 2 emissions, will only reinforce these plans. The ABWR will play an important role in meeting the conflicting needs of developed and developing economies for more massive amounts of electricity and the need worldwide to limit CO 2 emissions. Successful ABWR projects in Japan arid Taiwan, coupled with licensing approval in the United States, represent the new approach to the design, licensing, construction and operation of nuclear power in the new millennium. (author)

Self-Sustaining Thorium Boiling Water Reactors

Directory of Open Access Journals (Sweden)

Ehud Greenspan

2012-10-01

Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

Multi-core processing and scheduling performance in CMS

CERN Multimedia

CERN. Geneva

2012-01-01

Commodity hardware is going many-core. We might soon not be able to satisfy the job memory needs per core in the current single-core processing model in High Energy Physics. In addition, an ever increasing number of independent and incoherent jobs running on the same physical hardware not sharing resources might significantly affect processing performance. It will be essential to effectively utilize the multi-core architecture. CMS has incorporated support for multi-core processing in the event processing framework and the workload management system. Multi-core processing jobs share common data in memory, such us the code libraries, detector geometry and conditions data, resulting in a much lower memory usage than standard single-core independent jobs. Exploiting this new processing model requires a new model in computing resource allocation, departing from the standard single-core allocation for a job. The experiment job management system needs to have control over a larger quantum of resource since multi-...

Synthesis of Au@Ag core-shell nanocubes containing varying shaped cores and their localized surface plasmon resonances.

Science.gov (United States)

Gong, Jianxiao; Zhou, Fei; Li, Zhiyuan; Tang, Zhiyong

2012-06-19

We have synthesized Au@Ag core-shell nanocubes containing Au cores with varying shapes and sizes through modified seed-mediated methods. Bromide ions are found to be crucial in the epitaxial growth of Ag atoms onto Au cores and in the formation of the shell's cubic shape. The Au@Ag core-shell nanocubes exhibit very abundant and distinct localized surface plasmon resonance (LSPR) properties, which are core-shape and size-dependent. With the help of theoretical calculation, the physical origin and the resonance mode profile of each LSPR peak are identified and studied. The core-shell nanocrystals with varying shaped cores offer a new rich category for LSPR control through the plasmonic coupling effect between core and shell materials.

Crosshole investigations - physical properties of core samples from boreholes F1 and F2

International Nuclear Information System (INIS)

Magnusson, K.A.; Carlsten, S.; Olsson, O.

1987-06-01

The geology and physical properties have been studied of roughly 100 core samples from the boreholes F1 and F2 drilled at the Crosshole site, located at the 360 m level in the Stripa mine. The granitic rock has been divided into two classes: Fracture zones (also called major units) and a rock mass which is relatively undeformed. Samples from the major units have lower resistivity, higher porosity and dielectric constant than the samples from the less deformed rock mass. The electrical properties of the core samples have been measured over a frequency interval ranging from 1 Hz to 70 MHz. The conductivity of the samples increases with frequency, approximately with the frequency raised to the power 0.38. The dielectric constant decreases with frequency but is essentially constant above 3 MHz. These results show that the Hanai-Bruggeman equation can be used to describe the electrical bulk properties of the Stripa granite. The electrical conductivity of the samples is well correlated to the water content of the samples. The granite has a small contents of electrically conductive minerals which could influence the electrical bulk properties. (orig.)

Development of concept and neutronic calculation method for large LMFBR core

International Nuclear Information System (INIS)

Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.

1991-01-01

Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)

Surry Power Station, Units 1 and 2. Annual operating report for 1976, section 2: core performance and startup physics test reports

International Nuclear Information System (INIS)

1977-01-01

The analyses of four core performance indicators including burnup distribution, reactivity depletion, power distribution, and primary coolant activity are discussed. Information is presented concerning fuel densification monitoring; start-up physics testing; and James River temperature, salinity, fish, biota, and plankton entrainment monitoring

Developing a computational tool for predicting physical parameters of a typical VVER-1000 core based on artificial neural network

International Nuclear Information System (INIS)

Mirvakili, S.M.; Faghihi, F.; Khalafi, H.

2012-01-01

Highlights: ► Thermal–hydraulics parameters of a VVER-1000 core based on neural network (ANN), are carried out. ► Required data for ANN training are found based on modified COBRA-EN code and then linked each other using MATLAB software. ► Based on ANN method, average and maximum temperature of fuel and clad as well as MDNBR of each FA are predicted. -- Abstract: The main goal of the present article is to design a computational tool to predict physical parameters of the VVER-1000 nuclear reactor core based on artificial neural network (ANN), taking into account a detailed physical model of the fuel rods and coolant channels in a fuel assembly. Predictions of thermal characteristics of fuel, clad and coolant are performed using cascade feed forward ANN based on linear fission power distribution and power peaking factors of FAs and hot channels factors (which are found based on our previous neutronic calculations). A software package has been developed to prepare the required data for ANN training which applies a modified COBRA-EN code for sub-channel analysis and links the codes using the MATLAB software. Based on the current estimation system, five main core TH parameters are predicted, which include the average and maximum temperatures of fuel and clad as well as the minimum departure from nucleate boiling ratio (MDNBR) for each FA. To get the best conditions for the considered ANNs training, a comprehensive sensitivity study has been performed to examine the effects of variation of hidden neurons, hidden layers, transfer functions, and the learning algorithms on the training and simulation results. Performance evaluation results show that the developed ANN can be trained to estimate the core TH parameters of a typical VVER-1000 reactor quickly without loss of accuracy.

Reactor core performance estimating device

International Nuclear Information System (INIS)

Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

1995-01-01

The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

Energy Technology Data Exchange (ETDEWEB)

Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

2015-10-15

With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

International Nuclear Information System (INIS)

Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong

2015-01-01

With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

In-core fuel management for the course on operational physics of power reactors

International Nuclear Information System (INIS)

Levine, S.H.

1982-01-01

The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

Models of the earth's core

International Nuclear Information System (INIS)

Stevenson, D.J.

1981-01-01

The combination of seismology, high pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to strong constraints on core models. The synthesis presented here is devoted to the defense of the following properties: (1) core formation was contemporaneous with earth accretion; (2) the outer, liquid core is predominately iron but cannot be purely iron; (3) the inner core-outer core boundary represents a thermodynamic equilibrium between a liquid alloys and a predominanately iron solid; (4) thermodynamic and transport properties of outer core can be estimated from liquid-state theories; and (5) the outer core is probably adiabatic and uniform in composition. None of these propositions are universally accepted by geophysicists. But, the intent of this paper is to present a coherent picture which explains most of the data with the fewest ad hoc assumptions. Areas in which future progress is both essential and likely are geo- and cosmochronology, seismological determinations of core structure, fluid dynamics of the core and mantle, and condensed matter physics

Physical aspects of relaxation and shake-up effects in XPS and core →2π* absorption spectra of CO chemisorbed on Ni (111)

International Nuclear Information System (INIS)

Gumhalter, B.

1985-07-01

The physical origin of the peculiar relaxation shifts and spectral shapes appearing in x-ray induced core-to-valence excitation and core level photoemission spectra of CO chemisorbed on Ni(111) are discussed and interpreted within a unique framework. Within the model presented the electronic transitions in core-to-valence excitation spectroscopy and XPS are shown to give rise to drastic electronic rearrangements within the adsorption system and to the charge shake-up in the CO 2π* derived resonance partly filled via the backdonation mechanism. Such singular relaxation processes, common to both spectroscopies, are closely related and can be treated on the same footing. This makes possible to establish unique relaxation shifts and spectral characteristics for two seemingly different experimental situations. The use of this formalism in analysing the experimental data enables one to estimate and distinguish between the extra-adsorbate (image or nonbonding) and intra-adsorbate (chemically induced) screening of the core holes created either by x-ray induced core-to-valence electronic transitions or core level photoionization in CO/Ni(111). (author)

Analysis of the European results on the HTTR's core physics benchmarks

International Nuclear Information System (INIS)

Raepsaet, X.; Damian, F.; Ohlig, U.A.; Brockmann, H.J.; Haas, J.B.M. de; Wallerboss, E.M.

2002-01-01

Within the frame of the European contract HTR-N1 calculations are performed on the benchmark problems of the HTTR's start-up core physics experiments initially proposed by the IAEA in a Co-ordinated Research Programme. Three European partners, the FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France, have joined this work package with the aim to validate their calculational methods. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark. In the case of the diffusion codes, this discrepancy is reduced to: Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA). (author)

Multi-core MgO NPs(at)C core-shell nanospheres for selective CO2 capture under mild conditions

International Nuclear Information System (INIS)

Tae Kyung Kim; Kyung Joo Lee; Hoi Ri Moon; Junhan Yuh; Sang Kyu Kwak

2014-01-01

The core-shell structures have attracted attention in catalysis, because the outer shells isolate the catalytically active NP cores and prevent the possibility of sintering of core particles during catalytic reaction under physically and chemically harsh conditions. We aimed to adopt this core-shell system for CO 2 sorption materials. In this study, a composite material of multi-core 3 nm-sized magnesium oxide nanoparticles embedded in porous carbon nanospheres (MgO NPs(at)C) was synthesized by a gas phase reaction via a solvent-free process. It showed selective CO 2 adsorption capacity over N 2 under mild regeneration conditions. (authors)

The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

International Nuclear Information System (INIS)

1993-12-01

The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support

Core design and performance of small inherently safe LMRs

International Nuclear Information System (INIS)

Orechwa, Y.; Khalil, H.; Turski, R.B.; Fujita, E.K.

1986-01-01

Oxide and metal-fueled core designs at the 900 MWt level and constrained by a requirement for interchangeability are described. The physics parameters of the two cores studied here indicate that metal-fueled cores display attractive economic and safety features and are more flexible than are oxide cores in adapting to currently-changing deployment scenarios

Effects of core stabilization with and without conventional physical therapy for the management of non-specific low back pain

International Nuclear Information System (INIS)

Shoukat, F.; Ahmed, A.

2014-01-01

The purpose of the study was to examine the effects of core stabilization exercises with conventional physiotherapy for the management of non-specific low back pain (LBP). Methodology: This experimental comparative study was conducted at Department of Physiotherapy, PSRD hospital, Ferozpur Road Lahore. The study involved 40 subjects diagnosed with non-specific LBP with age ranges from 18 - 65 years. Patients were randomly allocated into 2 groups: treated with core stabilization exercises and conventional physiotherapy. Group - B (Control Group): In this group, patients were treated by conventional physiotherapy alone. The outcome measures were pain and physical functional outcomes. Pain was measured by using Visual Analo- gue Scale (V AS) and the physical functional outcomes of patients were measured by using Modified Oswes- tery Disability Questionnaire (MODQ). Results: By applying paired t-test in group - A, the p-values obtained for VAS and MODQ were statisti- cally significant (i.e., p = 0.000, p = 0.000 respectively) while in group - B, the p-values for VAS and MODQ were also found to be significant (p = 0.000, p = 0.000 respectively). By applying repeated measure analysis of variance (ANOVA), p-values were find to be insignificant for VAS (p = 0.09) and MODQ (p = 0.018). Conclusion: Both groups showed improvement in severity of pain and functional activity but the group- A that was given core stabilization exercises along with conventional treatment showed same improve- ment in pain and functional status as shown by group- B so any of the intervention can be used to gain better results as both are equally effective. (author)

Critical heat flux in a vertical annulus under low upward flow and near atmospheric pressure

International Nuclear Information System (INIS)

Schoesse, T.; Aritomi, Masanori; Lee, Sang-Ryoul; Kataoka, Yoshiaki; Yoshioka, Yuzuru; Chung, Moon-Ki.

1997-01-01

As future boiling water reactors (BWR), concepts of evolutional ABWR (ABWR-IER) and natural circulation BWR (JSBWR) have been investigated in order to reduce their construction cost and simplify their maintenance and inspection procedures. One of the promised features of the design of the evolutional ABWR is to reduce the number of internal pumps and to remove the Motor Generation (MG) sets. These design changes may induce boiling transition in the fuel rods of reactor core during a pump trip transient due to the more rapid flow coastdown characteristics than these of the present design. In addition, the understanding of critical heat flux (CHF) is one important subject to grasp safety margin during the start-up for the natural circulation BWR and to establish the rational start-up procedure in which thermo-hydraulic instabilities can be suppressed. The present study is to clarify CHF characteristics under low velocity conditions. CHF measurements were conducted in a vertical upward annulus channel composed of an inner heated rod and an outer tube made of glass. CHF data were obtained repeatedly under the condition of stable inlet flow to examine statistically their reproducibility. The flow regime was investigated from flow observation and measurement of differential pressure fluctuation. The CHF data are correlated with the flow regime transition. It was clear from the obtained flow pattern and the CHF data that the CHF behavior could be classified into specified regions by the mass flux and inlet subcooling conditions. A CHF correlation was developed and agreed with other researchers' data within acceptable error. (author)

Multi-core processing and scheduling performance in CMS

International Nuclear Information System (INIS)

Hernández, J M; Evans, D; Foulkes, S

2012-01-01

Commodity hardware is going many-core. We might soon not be able to satisfy the job memory needs per core in the current single-core processing model in High Energy Physics. In addition, an ever increasing number of independent and incoherent jobs running on the same physical hardware not sharing resources might significantly affect processing performance. It will be essential to effectively utilize the multi-core architecture. CMS has incorporated support for multi-core processing in the event processing framework and the workload management system. Multi-core processing jobs share common data in memory, such us the code libraries, detector geometry and conditions data, resulting in a much lower memory usage than standard single-core independent jobs. Exploiting this new processing model requires a new model in computing resource allocation, departing from the standard single-core allocation for a job. The experiment job management system needs to have control over a larger quantum of resource since multi-core aware jobs require the scheduling of multiples cores simultaneously. CMS is exploring the approach of using whole nodes as unit in the workload management system where all cores of a node are allocated to a multi-core job. Whole-node scheduling allows for optimization of the data/workflow management (e.g. I/O caching, local merging) but efficient utilization of all scheduled cores is challenging. Dedicated whole-node queues have been setup at all Tier-1 centers for exploring multi-core processing workflows in CMS. We present the evaluation of the performance scheduling and executing multi-core workflows in whole-node queues compared to the standard single-core processing workflows.

Sensitivity of physics parameters for establishment of a burned CANDU full-core model for decommissioning waste characterization

International Nuclear Information System (INIS)

Cho, Dong-Keun; Sun, Gwang-Min; Choi, Jongwon; Hwang, Dong-Hyun; Hwang, Tae-Won; Yang, Ho-Yeon; Park, Dong-Hwan

2011-01-01

The sensitivity of parameters related with reactor physics on the source terms of decommissioning wastes from a CANDU reactor was investigated in order to find a viable, simplified burned core model of a Monte Carlo simulation for decommissioning waste characterization. First, a sensitivity study was performed for the level of nuclide consideration in an irradiated fuel and implicit geometry modeling, the effects of side structural components of the core, and structural supporters for reactive devices. The overall effects for computation memory, calculation time, and accuracy were then investigated with a full-core model. From the results, it was revealed that the level of nuclide consideration and geometry homogenization are not important factors when the ratio of macroscopic neutron absorption cross section (MNAC) relative to a total value exceeded 0.95. The most important factor affecting the neutron flux of the pressure tube was shown to be the structural supporters for reactivity devices, showing an 10% difference. Finally, it was concluded that a bundle-average homogeneous model considering a MNAC of 0.95, which is the simplest model in this study, could be a viable approximate model, with about 25% lower computation memory, 40% faster simulation time, and reasonable engineering accuracy compared with a model with an explicit geometry employing an MNAC of 0.99. (author)

Nondestructive continuous physical property measurements of core samples recovered from hole B, Taiwan Chelungpu-Fault Drilling Project

Science.gov (United States)

Hirono, Tetsuro; Yeh, En-Chao; Lin, Weiren; Sone, Hiroki; Mishima, Toshiaki; Soh, Wonn; Hashimoto, Yoshitaka; Matsubayashi, Osamu; Aoike, Kan; Ito, Hisao; Kinoshita, Masataka; Murayama, Masafumi; Song, Sheng-Rong; Ma, Kuo-Fong; Hung, Jih-Hao; Wang, Chien-Ying; Tsai, Yi-Ben; Kondo, Tomomi; Nishimura, Masahiro; Moriya, Soichi; Tanaka, Tomoyuki; Fujiki, Toru; Maeda, Lena; Muraki, Hiroaki; Kuramoto, Toshikatsu; Sugiyama, Kazuhiro; Sugawara, Toshikatsu

2007-07-01

The Taiwan Chelungpu-Fault Drilling Project was undertaken in 2002 to investigate the faulting mechanism of the 1999 Mw 7.6 Taiwan Chi-Chi earthquake. Hole B penetrated the Chelungpu fault, and core samples were recovered from between 948.42- and 1352.60-m depth. Three major zones, designated FZB1136 (fault zone at 1136-m depth in hole B), FZB1194, and FZB1243, were recognized in the core samples as active fault zones within the Chelungpu fault. Nondestructive continuous physical property measurements, conducted on all core samples, revealed that the three major fault zones were characterized by low gamma ray attenuation (GRA) densities and high magnetic susceptibilities. Extensive fracturing and cracks within the fault zones and/or loss of atoms with high atomic number, but not a measurement artifact, might have caused the low GRA densities, whereas the high magnetic susceptibility values might have resulted from the formation of magnetic minerals from paramagnetic minerals by frictional heating. Minor fault zones were characterized by low GRA densities and no change in magnetic susceptibility, and the latter may indicate that these minor zones experienced relatively low frictional heating. Magnetic susceptibility in a fault zone may be key to the determination that frictional heating occurred during an earthquake on the fault.

GENOVA: a generalized perturbation theory program for various applications to CANDU core physics analysis (II) - a user's manual

International Nuclear Information System (INIS)

Kim, Do Heon; Choi, Hang Bok

2001-03-01

A user's guide for GENOVA, a GENeralized perturbation theory (GPT)-based Optimization and uncertainty analysis program for Canada deuterium uranium (CANDU) physics VAriables, was prepared. The program was developed under the framework of CANDU physics design and analysis code RFSP. The generalized perturbation method was implemented in GENOVA to estimate the zone controller unit (ZCU) level upon refueling operation and calculate various sensitivity coefficients for fuel management study and uncertainty analyses, respectively. This documentation contains descriptions and directions of four major modules of GENOVA such as ADJOINT, GADJINT, PERTURB, and PERTXS so that it can be used as a practical guide for GENOVA users. This documentation includes sample inputs for the ZCU level estimation and sensitivity coefficient calculation, which are the main application of GENOVA. The GENOVA can be used as a supplementary tool of the current CANDU physics design code for advanced CANDU core analysis and fuel development

Transition from 12 months to 18 months cycles at Krsko in a core physics perspective

International Nuclear Information System (INIS)

Jensen-Tornehed, J.

2004-01-01

Krsko has historically been operating in 12 months cycles with an annual outage for refuelling and maintenance work. Krsko is now in a transition from 12 months to 18 months cycles. Cycle 19, June 2002 - May 2003, was the last 12 month cycle. Cycle 20, June 2003 - September 2004 is a 15 month transition cycle and cycle 21, September 2004 - April 2006 will be the first 18 month cycle. This paper will describe the effects of the transition in a core physics perspective. There are big differences in how to design an 18 month cycle in comparison with a 12 month cycle. The required number of feed assemblies increases, as well as the content of burnable absorbers in the fuel. The strategy of the loading pattern has to be changed with the increased number of fresh fuel assemblies. The most limiting margins can be different for different cycle lengths which also affect the fresh assembly design and loading pattern during the transition. During the core design for cycle 21 the Moderator Temperature Coefficient was the main issue, which caused the need for extra amount of burnable absorbers. (author)

Modelling the core magnetic field of the earth

Science.gov (United States)

Harrison, C. G. A.; Carle, H. M.

1982-01-01

It is suggested that radial off-center dipoles located within the core of the earth be used instead of spherical harmonics of the magnetic potential in modeling the core magnetic field. The off-center dipoles, in addition to more realistically modeling the physical current systems within the core, are if located deep within the core more effective at removing long wavelength signals of either potential or field. Their disadvantage is that their positions and strengths are more difficult to compute, and such effects as upward and downward continuation are more difficult to manipulate. It is nevertheless agreed with Cox (1975) and Alldredge and Hurwitz (1964) that physical realism in models is more important than mathematical convenience. A radial dipole model is presented which agrees with observations of secular variation and excursions.

Strategy of plant concept development in Hitachi

International Nuclear Information System (INIS)

Kumiaki, Moriya

2007-01-01

Hitachi contributes to the society in various fields of nuclear power such as the Light Water Reactor field, the Fast Breeder Reactor, the fuel cycle, and the medical treatment. Since the beginning of a first commercial operation of a BWR in Japan, Hitachi has constructed 20 units of BWR. Hitachi continues its efforts in achieving high reliability and large-scale output, and in 1996, it completed Advanced Boiling Water Reactor (ABWR) in cooperation with various BWR utilities, General Electric Company, and Toshiba Corporation. Hitachi has enhanced the ABWR technology further based on the above enough experience. The latest technologies were reflected in Hamaoka unit 5 and Shika unit 2 as the latest ABWR plants. The further upgrade technologies would been reflected in Shimane-3, Ohma-1 and Higashidori-1 as ABWR plants under planning. Hitachi obtains the chance to construct the nuclear power plant continuously. For the next generation, Hitachi is working on developing nuclear power plants that take diversified needs and global characteristics into account. As one of the approaches, the output series formation of BWR is extended as the following. ABWR-II and ESBWR (Economic and Simplified BWR) as a large-scale centralized power supply emphasizing cost efficiency, Medium size ABWR and natural circulation type BWR as medium and small-scale distributed power supply that features flexibility to various market needs, such as minimized capital risks, timely return on a capital investment, etc. As another approach, Hitachi tries to extend the Light Water Reactor technology. RBWR (Resource-Renewable BWR ) that achieves a high conversion ratio over 1.0 based on the BWR technology will make the fuel cycle flexible. Hitachi will continue the challenge for the next ABWR and the future with the enough experience of BWR construction. (author)

GCRA review and appraisal of HTGR reactor-core-design program

International Nuclear Information System (INIS)

1980-09-01

The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

International Nuclear Information System (INIS)

1994-07-01

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

International Nuclear Information System (INIS)

1994-07-01

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

International Nuclear Information System (INIS)

Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

1993-12-01

Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

Automated Design and Optimization of Pebble-bed Reactor Cores

International Nuclear Information System (INIS)

Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

2010-01-01

We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

[Comparative investigation of compressive resistance of glass-cermet cements used as a core material in post-core systems].

Science.gov (United States)

Ersoy, E; Cetiner, S; Koçak, F

1989-09-01

In post-core applications, addition to the cast designs restorations that are performed on fabrication posts with restorative materials are being used. To improve the physical properties of glass-ionomer cements that are popular today, glass-cermet cements have been introduced and those materials have been proposed to be an alternative restorative material in post-core applications. In this study, the compressive resistance of Ketac-Silver as a core material was investigated comparatively with amalgam and composite resins.

Dehydration of core/shell fruits

OpenAIRE

Liu, Y.; Yang, Xiaosong; Cao, Y.; Wang, Z.; Chen, B.; Zhang, Jian J.; Zhang, H.

2015-01-01

Dehydrated core/shell fruits, such as jujubes, raisins and plums, show very complex buckles and wrinkles on their exocarp. It is a challenging task to model such complicated patterns and their evolution in a virtual environment even for professional animators. This paper presents a unified physically-based approach to simulate the morphological transformation for the core/shell fruits in the dehydration process. A finite element method (FEM), which is based on the multiplicative decomposition...

Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap; Khoi dong vat ly lo phan ung hat nhan Da Lat voi cau hinh vung hoat co bay notron

Energy Technology Data Exchange (ETDEWEB)

Hien, Pham Duy; Huy, Ngo Quang; Long, Vu Hai; Mai, Tran Khanh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs.

Applications of liquid state physics to the earth's core

Science.gov (United States)

Stevenson, D. J.

1980-01-01

New results derived for application to the earth's outer core using the modern theory of liquids and the hard-sphere model of liquid structure are presented. An expression derived in terms of the incompressibility and pressure is valid for a high-pressure liquid near its melting point, provided that the pressure is derived from a strongly repulsive pair potential; a relation derived between the melting point and density leads to a melting curve law of essentially the same form as Lindemann's law. Finally, it is shown that the 'core paradox' of Higgins and Kennedy (1971) can occur only if the Gruneisen parameter is smaller than 2/3, and this constant is larger than this value in any liquid for which the pair potential is strongly repulsive.

Trace analysis of loss of feedwater flow event in Lungmen ABWR

International Nuclear Information System (INIS)

Wang Jongrong; Lin Haotzu; Wang Weichen; Yang Shuming; Shih Chunkuan

2009-01-01

TRACE (TRAC/RELAP Advanced Computational Engine) model of Lungmen Nuclear Power Plant was used to analyze the Loss of Feedwater Flow transient as defined in Lungmen FSAR Chapter 15. The results were compared with those from FSAR and RETRAN02. Lungmen TRACE model will have two models: In model A, vessel is divided into 11 axial levels, 4 radial rings and 1 azimuthal sectors; In model B, vessel is divided into 11 axial levels, 4 radial rings, and 6 azimuthal sectors. The above models include feedwater control system, narrow range water level control system, and wide range water level control system. The loss of feedwater flow (LOFW) transient began with the trip of two operating feedwater pumps either from the pump mechanical/electric failure, or the operator human error, or high water level signal. Feedwater flow was assumed to descend to 0 in 5 seconds and led to the decrease of reactor water level. At L3 low water level setpoint, the system actuated reactor scram signal and RIP trip signal for RIPs not connected to the M/G set. At L2 low-low water level setpoint, the system would trip the other six RIPs. This paper compares those important thermal parameters at steady state, such as the dome pressure and temperature of reactor vessel, steam flow, feedwater flow, core flow, and RIP flow, etc.. It also compares system parameters under transient conditions, such as core thermal power, core flow, steam flow, feedwater flow, Narrow Range Water Level (NRWL), Wide Range Water Level (WRWL) and RIP flow, etc.. It was concluded that the steady state and transient results of TRACE calculations are in good agreement with those from RETRAN02. In summary, our studies concluded that Lungmen TRACE model is correct and accurate enough for future safety analysis applications. (author)

SMART core protection system design

International Nuclear Information System (INIS)

Lee, J. K.; Park, H. Y.; Koo, I. S.; Park, H. S.; Kim, J. S.; Son, C. H.

2003-01-01

SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system

Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core

International Nuclear Information System (INIS)

Ovdiienko, I.; Kuchin, A.; Khalimonchuk, V.; Ieremenko, M.

2011-01-01

Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)

Future directions in boiling water reactor design

International Nuclear Information System (INIS)

Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

1987-01-01

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

Effects of electron-withdrawing group and electron-donating core combinations on physical properties and photovoltaic performance in D-pi-A star-shaped small molecules

NARCIS (Netherlands)

Luponosov, Yuriy N.; Min, Jie; Solodukhin, Alexander N.; Kozlov, Oleg V.; Obrezkova, Marina A.; Peregudova, Svetlana M.; Ameri, Tayebeh; Chvalun, Sergei N.; Pshenichnikov, Maxim S.; Brabec, Christoph J.; Ponomarenko, Sergei A.

The first representatives of star-shaped molecules having 3-alkylrhodanine (alkyl-Rh) electron-withdrawing groups, linked through bithiophene pi-spacer with electron-donating either triphenylamine (TPA) or tris(2-methoxyphenyl)amine (m-TPA) core were synthesized. The physical properties and

NH3 (10-00) in the pre-stellar core L1544

DEFF Research Database (Denmark)

Caselli, P.; Bizzocchi, L.; Keto, E.

2017-01-01

GHz and study the abundance profile of ammonia across the pre-stellar core L1544 to test current theories of its physical and chemical structure. Recently calculated collisional coefficients have been included in our non-LTE radiative transfer code to reproduce Herschel observations. A gas......Pre-stellar cores represent the initial conditions in the process of star and planet formation, therefore it is important to study their physical and chemical structure. Because of their volatility, nitrogen-bearing molecules are key to study the dense and cold gas present in pre-stellar cores....... The NH3 rotational transition detected with Herschel-HIFI provides a unique combination of sensitivity and spectral resolution to further investigate physical and chemical processes in pre-stellar cores. Here we present the velocity-resolved Herschel-HIFI observations of the ortho-NH3(10-00) line at 572...

Feasibility analysis of real-time physical modeling using WaveCore processor technology on FPGA

NARCIS (Netherlands)

Verstraelen, Martinus Johannes Wilhelmina; Pfeifle, Florian; Bader, Rolf

2015-01-01

WaveCore is a scalable many-core processor technology. This technology is specifically developed and optimized for real-time acoustical modeling applications. The programmable WaveCore soft-core processor is silicon-technology independent and hence can be targeted to ASIC or FPGA technologies. The

Positive experience in the construction and project management of Kashiwazaki-Kariwa no. 6 and no. 7

International Nuclear Information System (INIS)

Minematsu, A.

1999-01-01

Kashiwazaki-Kariwa No.6 and No.7 (K-6/7), the world first Advanced Boiling Water Reactor (ABWR) units, started commercial operations on November 7, 1996 and July 2, 1997 respectively. ABWR has been developed as a standard design of the next generation BWR to meet common goals set by the Japanese electric utilities and BWR manufacturers (GE, Hitachi and Toshiba) based on our design, construction, operation and maintenance experience of nuclear power plants. The construction of K-6/7 and confirmatory tests for the verification of the first-of-a-kind (FOAK) design features of ABWR were conducted smoothly without any delay. The duration of the construction was 51.5 months. It was shorter than conventional BWRs in Japan by nearly one year. This was realized by design features of ABWR for better constructability, a principle of 'test before use' applied to the FOAKs, advanced construction technology, detailed engineering at very early stages of the project, and good construction management. The positive experience in the K-6/7 project is now being reviewed and standardized for next ABWR projects. The data and knowledge accumulated through the K-6/7 project will be utilized effectively with the aide of the latest information technology. (author)

Physics Meets Art in the General Education Core

Science.gov (United States)

Dark, Marta L.; Hylton, Derrick J.

2018-01-01

This article describes a general education course offering, Physics and the Arts. During the development of this course, physics and arts faculty collaborated closely. We cover the usual physics phenomena for such a course--light, color, and sound--in addition to gravity, equilibrium, and spacetime. Goals of the course are to increase students'…

APROS 3-D core models for simulators and plant analyzers

International Nuclear Information System (INIS)

Puska, E.K.

1999-01-01

The 3-D core models of APROS simulation environment can be used in simulator and plant analyzer applications, as well as in safety analysis. The key feature of APROS models is that the same physical models can be used in all applications. For three-dimensional reactor cores the APROS models cover both quadratic BWR and PWR cores and the hexagonal lattice VVER-type cores. In APROS environment the user can select the number of flow channels in the core and either five- or six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the channel description have a decisive effect on the calculation time of the 3-D core model and thus just these selection make at present the major difference between a safety analysis model and a training simulator model. The paper presents examples of various types of 3-D LWR-type core descriptions for simulator and plant analyzer use and discusses the differences of calculation speed and physical results between a typical safety analysis model description and a real-time simulator model description in transients. (author)

High-resolution gamma ray attenuation density measurements on mining exploration drill cores, including cut cores

Science.gov (United States)

Ross, P.-S.; Bourke, A.

2017-01-01

Physical property measurements are increasingly important in mining exploration. For density determinations on rocks, one method applicable on exploration drill cores relies on gamma ray attenuation. This non-destructive method is ideal because each measurement takes only 10 s, making it suitable for high-resolution logging. However calibration has been problematic. In this paper we present new empirical, site-specific correction equations for whole NQ and BQ cores. The corrections force back the gamma densities to the "true" values established by the immersion method. For the NQ core caliber, the density range extends to high values (massive pyrite, 5 g/cm3) and the correction is thought to be very robust. We also present additional empirical correction factors for cut cores which take into account the missing material. These "cut core correction factors", which are not site-specific, were established by making gamma density measurements on truncated aluminum cylinders of various residual thicknesses. Finally we show two examples of application for the Abitibi Greenstone Belt in Canada. The gamma ray attenuation measurement system is part of a multi-sensor core logger which also determines magnetic susceptibility, geochemistry and mineralogy on rock cores, and performs line-scan imaging.

Seismic response of a block-type nuclear reactor core

International Nuclear Information System (INIS)

Dove, R.C.; Bennett, J.G.; Merson, J.L.

1976-05-01

An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

Core-Collapse Supernovae, Neutrinos, and Gravitational Waves

Energy Technology Data Exchange (ETDEWEB)

Ott, C.D. [TAPIR, California Institute of Technology, Pasadena, California (United States); Kavli Institute for the Physics and Mathematics of the Universe, Kashiwa, Chiba (Japan); O' Connor, E.P. [Canadian Institute for Theoretical Astrophysics, Toronto, Ontario (Canada); Gossan, S.; Abdikamalov, E.; Gamma, U.C.T. [TAPIR, California Institute of Technology, Pasadena, California (United States); Drasco, S. [Grinnell College, Grinnell, Iowa (United States); TAPIR, California Institute of Technology, Pasadena, California (United States)

2013-02-15

Core-collapse supernovae are among the most energetic cosmic cataclysms. They are prodigious emitters of neutrinos and quite likely strong galactic sources of gravitational waves. Observation of both neutrinos and gravitational waves from the next galactic or near extragalactic core-collapse supernova will yield a wealth of information on the explosion mechanism, but also on the structure and angular momentum of the progenitor star, and on aspects of fundamental physics such as the equation of state of nuclear matter at high densities and low entropies. In this contribution to the proceedings of the Neutrino 2012 conference, we summarize recent progress made in the theoretical understanding and modeling of core-collapse supernovae. In this, our emphasis is on multi-dimensional processes involved in the explosion mechanism such as neutrino-driven convection and the standing accretion shock instability. As an example of how supernova neutrinos can be used to probe fundamental physics, we discuss how the rise time of the electron antineutrino flux observed in detectors can be used to probe the neutrino mass hierarchy. Finally, we lay out aspects of the neutrino and gravitational-wave signature of core-collapse supernovae and discuss the power of combined analysis of neutrino and gravitational wave data from the next galactic core-collapse supernova.

Core-Collapse Supernovae, Neutrinos, and Gravitational Waves

International Nuclear Information System (INIS)

Ott, C.D.; O'Connor, E.P.; Gossan, S.; Abdikamalov, E.; Gamma, U.C.T.; Drasco, S.

2013-01-01

Core-collapse supernovae are among the most energetic cosmic cataclysms. They are prodigious emitters of neutrinos and quite likely strong galactic sources of gravitational waves. Observation of both neutrinos and gravitational waves from the next galactic or near extragalactic core-collapse supernova will yield a wealth of information on the explosion mechanism, but also on the structure and angular momentum of the progenitor star, and on aspects of fundamental physics such as the equation of state of nuclear matter at high densities and low entropies. In this contribution to the proceedings of the Neutrino 2012 conference, we summarize recent progress made in the theoretical understanding and modeling of core-collapse supernovae. In this, our emphasis is on multi-dimensional processes involved in the explosion mechanism such as neutrino-driven convection and the standing accretion shock instability. As an example of how supernova neutrinos can be used to probe fundamental physics, we discuss how the rise time of the electron antineutrino flux observed in detectors can be used to probe the neutrino mass hierarchy. Finally, we lay out aspects of the neutrino and gravitational-wave signature of core-collapse supernovae and discuss the power of combined analysis of neutrino and gravitational wave data from the next galactic core-collapse supernova

Understanding Core-Collapse Supernovae

Science.gov (United States)

Hix, W. R.; Lentz, E. J.; Baird, M.; Messer, O. E. B.; Mezzacappa, A.; Lee, C.-T.; Bruenn, S. W.; Blondin, J. M.; Marronetti, P.

2010-03-01

Our understanding of core-collapse supernovae continues to improve as better microphysics is included in increasingly realistic neutrino-radiationhydrodynamic simulations. Recent multi-dimensional models with spectral neutrino transport, which slowly develop successful explosions for a range of progenitors between 12 and 25 solar mass, have motivated changes in our understanding of the neutrino reheating mechanism. In a similar fashion, improvements in nuclear physics, most notably explorations of weak interactions on nuclei and the nuclear equation of state, continue to refine our understanding of how supernovae explode. Recent progresses on both the macroscopic and microscopic effects that affect core-collapse supernovae are discussed.

Start up physics tests of units 5 and 6 (WWER 1000) at Kozloduy NPP by comparison with the calculated neutron physics characteristics

International Nuclear Information System (INIS)

Antov, A.; Stoyanova, I.

2008-01-01

In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests, which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies (FAs). The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6. (authors)

Start up physics tests of Units 5 and 6 (WWER-1000) at Kozloduy NPP by comparison with the calculated neutron physics characteristics

International Nuclear Information System (INIS)

Stoyanova, I.; Antov, A.

2007-01-01

In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies. The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6 (Authors)

Core-in-shell sorbent for hot coal gas desulfurization

Science.gov (United States)

Wheelock, Thomas D.; Akiti, Jr., Tetteh T.

2004-02-10

A core-in-shell sorbent is described herein. The core is reactive to the compounds of interest, and is preferably calcium-based, such as limestone for hot gas desulfurization. The shell is a porous protective layer, preferably inert, which allows the reactive core to remove the desired compounds while maintaining the desired physical characteristics to withstand the conditions of use.

Advanced gadolinia core and Toshiba advanced reactor management system

International Nuclear Information System (INIS)

Miyamoto, Toshiki; Yoshioka, Ritsuo; Ebisuya, Mitsuo

1988-01-01

At the Hamaoka Nuclear Power Station, Unit No. 3, advanced core design and core management technology have been adopted, significantly improving plant availability, operability and reliability. The outstanding technologies are the advanced gadolinia core (AGC) which utilizes gadolinium for the axial power distribution control, and Toshiba advanced reactor management system (TARMS) which uses a three-dimensional core physics simulator to calculate the power distribution. Presented here are the effects of these advanced technologies as observed during field testing. (author)

Photo-physical properties enhancement of bare and core-shell quantum dots

Energy Technology Data Exchange (ETDEWEB)

Mumin, Md Abdul, E-mail: pcharpentier@eng.uwo.ca; Akhter, Kazi Farida, E-mail: pcharpentier@eng.uwo.ca; Charpentier, Paul A., E-mail: pcharpentier@eng.uwo.ca [Chemical and Biochemical Engineering, Western University, London Ontario (Canada)

2014-03-31

Semiconductor nanocrystals (NCs) (also known as quantum dots, QDs) have attracted immense attention for their size-tunable optical properties that makes them impressive candidates for solar cells, light emitting devices, lasers, as well as biomedical imaging. However monodispersity, high and consistent photoluminescence, photostability, and biocompatibility are still major challenges. This work focuses on optimizing the photophysical properties and biocompatibility of QDs by forming core-shell nanostructures and their encapsulation by a carrier. Highly luminescent CdS and CdS-ZnS core-shell QDs with 5 nm sizes were synthesized using a facile approach based on pyrolysis of the single molecule precursors. After capping the CdS QDs with a thin layer of ZnS to reduce toxicity, the photoluminescence and photostability of the core-shell QDs was significantly enhanced. To make both the bare and core/shell structure QDs more resistant against photochemical reactions, a mesoporous silica layer was grown on the QDs through a reverse microemulsion technique based on hydrophobic interaction. This encapsulation enhanced the quantum yield and photostability compared to the bare QDs by providing much stronger resistance to oxidation and Oswald ripening of QDs. Encapsulation also improved biocompatibility of QDs that was evaluated with human umbilical vein endothelial cell lines (HUVEC)

MA-core loaded untuned RF compression cavity for HIRFL-CSR

International Nuclear Information System (INIS)

Mei Lirong; Xu Zhe; Yuan Youjin; Jin Peng; Bian Zhibin; Zhao Hongwei; Xia Jiawen

2012-01-01

To meet the requirements of high energy density physics and plasma physics research at HIRFL-CSR the goal of achieving a higher accelerating gap voltage was proposed. Therefore, a magnetic alloy (MA)-core loaded radio frequency (RF) cavity that can provide a higher accelerating gap voltage compared to standard ferrite loaded cavities has been studied at IMP. In order to select the proper magnetic alloy material to load the RF compression cavity, measurements of four different kinds of sample MA-cores have been carried out. By testing the small cores, the core composition was selected to obtain the desired performance. According to the theoretical calculation and simulation, which show reasonable consistency for the MA-core loaded cavity, the desired performance can be achieved. Finally about 1000 kW power will be needed to meet the requirements of 50 kV accelerating gap voltage by calculation.

Development of the evaluation methods in reactor safety analyses and core characteristics

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

Nuclear design and analysis report for KALIMER breakeven core conceptual design

International Nuclear Information System (INIS)

Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

2002-04-01

During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

Random close packing in protein cores.

Science.gov (United States)

Gaines, Jennifer C; Smith, W Wendell; Regan, Lynne; O'Hern, Corey S

2016-03-01

Shortly after the determination of the first protein x-ray crystal structures, researchers analyzed their cores and reported packing fractions ϕ ≈ 0.75, a value that is similar to close packing of equal-sized spheres. A limitation of these analyses was the use of extended atom models, rather than the more physically accurate explicit hydrogen model. The validity of the explicit hydrogen model was proved in our previous studies by its ability to predict the side chain dihedral angle distributions observed in proteins. In contrast, the extended atom model is not able to recapitulate the side chain dihedral angle distributions, and gives rise to large atomic clashes at side chain dihedral angle combinations that are highly probable in protein crystal structures. Here, we employ the explicit hydrogen model to calculate the packing fraction of the cores of over 200 high-resolution protein structures. We find that these protein cores have ϕ ≈ 0.56, which is similar to results obtained from simulations of random packings of individual amino acids. This result provides a deeper understanding of the physical basis of protein structure that will enable predictions of the effects of amino acid mutations to protein cores and interfaces of known structure.

Physics of high-temperature reactors

International Nuclear Information System (INIS)

Massimo, L.

1976-01-01

The subject is covered in chapters entitled: general description of the HTR core; general considerations about reactor physics; neutron cross-sections; basic aspects of transport and diffusion theory; methods for the solution of the diffusion equation; slowing-down and thermalization in graphite; resonance absorption; spectrum calculations and cross-section averaging; burn-up; core design; fuel management and cost calculations; temperature coefficient; core dynamics and accident analysis; reactor control; peculiarities of HTR physics; analysis of calculational accuracy; sequence of reactor design calculations. (U.K.)

Modeling of the reactor core

International Nuclear Information System (INIS)

1999-01-01

In order to improve technical - economical parameters fuel with 2.4% enrichment and burnable absorber is started to be used at Ignalina NPP. Using code QUABOX/CUBBOX the main neutronic - physical characteristics were calculated for selected reactor core conditions

Introduction of characteristics and construction methods for Lungmen NPP

International Nuclear Information System (INIS)

Hur, Kyung Phil; Manager, General

2007-01-01

Daewoo Engineering and Construction Co., Ltd. has taken part in the construction of ABWR (Advance Boiling Water Reactor of capacity of 1,350MW) at Lungmen site in Taiwan since 1999 as a member of Joint Company for the Nuclear Island Civil Work and Nuclear Island Equipment and Piping Work. This paper discusses the introduction of NSSS of ABWR and the major construction methods applied to Lungmen Nuclear Power Plants. Even though there is no BWR or ABWR type of Nuclear Power Plants built in Korea, the information and data related to the newly advanced construction technologies which has been already applied successfully is very important for the future overseas expansion

Outline of advanced boiling water reactor

International Nuclear Information System (INIS)

Yoshio Matsuo

1987-01-01

The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

Physical domains in plasma physics

International Nuclear Information System (INIS)

Liboff, R.L.

1987-01-01

Do the plasma in the sun's core and the electron-conduction plasma in a semiconductor behave in the same way? This question is both fundamental and practical, for plasma physics plays a role in a vast area of natural phenomena and in many engineering devices. Understanding the cosmos, or designing a computer chip or a thermonuclear fusion reactor, requires first of all a realization of equations of motion that are appropriate to the particular problem. Similar physical differences occur in engineered structures. The plasmas in most thermonuclear fusion devices are basically like the plasma in the core of the sun: weakly coupled and classical - that is, obeying Newton's laws and Maxwell's equations. The conduction electrons in a semiconductor, on the other hand, obey the laws of quantum mechanics

MDEP Design-Specific Common Position CP-ABWRWG-01. Common position addressing issues related to the Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

2016-01-01

On 11 March 2011 Japan suffered its worst recorded earthquake. The epicentre was 110 miles (180 km) east north east from TEPCO's Fukushima Daiichi NPP which had 6 operating Boiling Water Reactors (BWR). Reactor Units 1, 2 and 3 on this site were operating at power before the event and on detection of the earthquake shut down safely. Reactor Unit 4 was not loaded with any fuel assemblies due to maintenance, including replacement of the core shroud. Reactor Units 5 and 6 were shut down for maintenance. Initially, on-site back-up diesel generators were used to provide the alternating current (AC) electrical supplies to power essential post-trip cooling systems. Within an hour a massive tsunami from the earthquake inundated the site; all AC electrical power to the cooling systems for the reactor and reactor fuel pools was lost, including that from all but one (in Unit 6) back-up diesel generators, as well as a significant amount of direct current (DC) supplies and essential instrumentation. Over the next few hours, and days, the fuel in Reactor Units 1, 2 and 3 heated up and degraded. There were hydrogen explosions which caused considerable damage to the reactor buildings in Units 1, 3 and 4. For over a week the site struggled to put cooling water into the reactors and the reactor fuel pools, by using means untried before and/or unplanned. Electrical supplies were gradually reconnected to the reactor buildings and a degree of control returned. Heavily contaminated water, used to cool the reactors and spent fuel pools, collected in un-contained areas of the site and leaked out to sea. There were also significant releases of radioactivity to the air. However, to date, the indications are that the public health effects from radiation exposure are not significant. This was a serious nuclear accident, with an International Nuclear and Radiological Event Scale (INES) rating of Level 7 (the highest level). Tens of thousands of people were evacuated from a zone extending

Magnetic core-shell silica particles

NARCIS (Netherlands)

Claesson, E.M.

2007-01-01

This thesis deals with magnetic silica core-shell colloids and related functionalized silica structures. Synthesis routes have been developed and optimized. The physical properties of these colloids have been investigated, such as the magnetic dipole moment, dipolar structure formation and

TMI-2 core debris grab samples: Examination and analysis: Part 1

International Nuclear Information System (INIS)

Akers, D.W.; Carlson, E.R.; Cook, B.A.; Ploger, S.A.; Carlson, J.O.

1986-09-01

Six samples of particulate debris were removed from the TMI-2 core rubble bed during September and October 1983, and five more samples were obtained in March 1984. The samples (up to 174 g each) were obtained at two locations in the core: H8 (center) and E9 (mid-radius). Ten of the eleven samples were examined at the Idaho National Engineering Laboratory to obtain data on the physical and chemical nature of the debris and the postaccident condition of the core. Portions of the samples also were subjected to differential thermal analysis at Rockwell Hanford Operations and metallurgical and chemical examinations at Argonne National Laboratories. This report presents results of the examination of the core debris grab samples, including physical, metallurgical, chemical, and radiochemical analyses. The results indicate that temperatures in the core reached at least 3100 K during the TMI-2 accident, fuel melting and significant mixing of core structural material occurred, and large fractions of some radionuclides (e.g., 90 Sr and 144 Ce) were retained in the core

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

International Nuclear Information System (INIS)

Wagner, K.C.

1988-10-01

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

Multi-objective and multi-physics optimization methodology for SFR core: application to CFV concept

International Nuclear Information System (INIS)

Fabbris, Olivier

2014-01-01

Nuclear reactor core design is a highly multidisciplinary task where neutronics, thermal-hydraulics, fuel thermo-mechanics and fuel cycle are involved. The problem is moreover multi-objective (several performances) and highly dimensional (several tens of design parameters).As the reference deterministic calculation codes for core characterization require important computing resources, the classical design method is not well suited to investigate and optimize new innovative core concepts. To cope with these difficulties, a new methodology has been developed in this thesis. Our work is based on the development and validation of simplified neutronics and thermal-hydraulics calculation schemes allowing the full characterization of Sodium-cooled Fast Reactor core regarding both neutronics performances and behavior during thermal hydraulic dimensioning transients.The developed methodology uses surrogate models (or meta-models) able to replace the neutronics and thermal-hydraulics calculation chain. Advanced mathematical methods for the design of experiment, building and validation of meta-models allows substituting this calculation chain by regression models with high prediction capabilities.The methodology is applied on a very large design space to a challenging core called CFV (French acronym for low void effect core) with a large gain on the sodium void effect. Global sensitivity analysis leads to identify the significant design parameters on the core design and its behavior during unprotected transient which can lead to severe accidents. Multi-objective optimizations lead to alternative core configurations with significantly improved performances. Validation results demonstrate the relevance of the methodology at the pre-design stage of a Sodium-cooled Fast Reactor core. (author) [fr

Magnetohydrodynamic Convection in the Outer Core and its Geodynamic Consequences

Science.gov (United States)

Kuang, Weijia; Chao, Benjamin F.; Fang, Ming

2004-01-01

The Earth's fluid outer core is in vigorous convection through much of the Earth's history. In addition to generating and maintaining Earth s time-varying magnetic field (geodynamo), the core convection also generates mass redistribution in the core and a dynamical pressure field on the core-mantle boundary (CMB). All these shall result in various core-mantle interactions, and contribute to surface geodynamic observables. For example, electromagnetic core-mantle coupling arises from finite electrically conducting lower mantle; gravitational interaction occurs between the cores and the heterogeneous mantle; mechanical coupling may also occur when the CMB topography is aspherical. Besides changing the mantle rotation via the coupling torques, the mass-redistribution in the core shall produce a spatial-temporal gravity anomaly. Numerical modeling of the core dynamical processes contributes in several geophysical disciplines. It helps explain the physical causes of surface geodynamic observables via space geodetic techniques and other means, e.g. Earth's rotation variation on decadal time scales, and secular time-variable gravity. Conversely, identification of the sources of the observables can provide additional insights on the dynamics of the fluid core, leading to better constraints on the physics in the numerical modeling. In the past few years, our core dynamics modeling efforts, with respect to our MoSST model, have made significant progress in understanding individual geophysical consequences. However, integrated studies are desirable, not only because of more mature numerical core dynamics models, but also because of inter-correlation among the geophysical phenomena, e.g. mass redistribution in the outer core produces not only time-variable gravity, but also gravitational core-mantle coupling and thus the Earth's rotation variation. They are expected to further facilitate multidisciplinary studies of core dynamics and interactions of the core with other

Core List of Astronomy and Physics Journals

Science.gov (United States)

Bryson, Liz; Fortner, Diane; Yorks, Pamela

This is a list of highly-used and highly-cited physics and astronomy journals. "Use" is measured largely on paper-journal counts from selective academic research-level libraries. Citation count titles are drawn from Institute for Scientific Information (ISI) data. Recognition is given to entrepreneurial electronic-only or new-style electronic journals. Selective news, magazine, and general science journals are omitted. The compilers welcome questions, suggestions for additions, or other advice. Comments may be sent c/o Diane Fortner, Physics Library, University of California, Berkeley. Dfortner@library.berkeley.edu

Exploring cosmic origins with CORE: Cosmological parameters

Science.gov (United States)

Di Valentino, E.; Brinckmann, T.; Gerbino, M.; Poulin, V.; Bouchet, F. R.; Lesgourgues, J.; Melchiorri, A.; Chluba, J.; Clesse, S.; Delabrouille, J.; Dvorkin, C.; Forastieri, F.; Galli, S.; Hooper, D. C.; Lattanzi, M.; Martins, C. J. A. P.; Salvati, L.; Cabass, G.; Caputo, A.; Giusarma, E.; Hivon, E.; Natoli, P.; Pagano, L.; Paradiso, S.; Rubiño-Martin, J. A.; Achúcarro, A.; Ade, P.; Allison, R.; Arroja, F.; Ashdown, M.; Ballardini, M.; Banday, A. J.; Banerji, R.; Bartolo, N.; Bartlett, J. G.; Basak, S.; Baumann, D.; de Bernardis, P.; Bersanelli, M.; Bonaldi, A.; Bonato, M.; Borrill, J.; Boulanger, F.; Bucher, M.; Burigana, C.; Buzzelli, A.; Cai, Z.-Y.; Calvo, M.; Carvalho, C. S.; Castellano, G.; Challinor, A.; Charles, I.; Colantoni, I.; Coppolecchia, A.; Crook, M.; D'Alessandro, G.; De Petris, M.; De Zotti, G.; Diego, J. M.; Errard, J.; Feeney, S.; Fernandez-Cobos, R.; Ferraro, S.; Finelli, F.; de Gasperis, G.; Génova-Santos, R. T.; González-Nuevo, J.; Grandis, S.; Greenslade, J.; Hagstotz, S.; Hanany, S.; Handley, W.; Hazra, D. K.; Hernández-Monteagudo, C.; Hervias-Caimapo, C.; Hills, M.; Kiiveri, K.; Kisner, T.; Kitching, T.; Kunz, M.; Kurki-Suonio, H.; Lamagna, L.; Lasenby, A.; Lewis, A.; Liguori, M.; Lindholm, V.; Lopez-Caniego, M.; Luzzi, G.; Maffei, B.; Martin, S.; Martinez-Gonzalez, E.; Masi, S.; Matarrese, S.; McCarthy, D.; Melin, J.-B.; Mohr, J. J.; Molinari, D.; Monfardini, A.; Negrello, M.; Notari, A.; Paiella, A.; Paoletti, D.; Patanchon, G.; Piacentini, F.; Piat, M.; Pisano, G.; Polastri, L.; Polenta, G.; Pollo, A.; Quartin, M.; Remazeilles, M.; Roman, M.; Ringeval, C.; Tartari, A.; Tomasi, M.; Tramonte, D.; Trappe, N.; Trombetti, T.; Tucker, C.; Väliviita, J.; van de Weygaert, R.; Van Tent, B.; Vennin, V.; Vermeulen, G.; Vielva, P.; Vittorio, N.; Young, K.; Zannoni, M.

2018-04-01

We forecast the main cosmological parameter constraints achievable with the CORE space mission which is dedicated to mapping the polarisation of the Cosmic Microwave Background (CMB). CORE was recently submitted in response to ESA's fifth call for medium-sized mission proposals (M5). Here we report the results from our pre-submission study of the impact of various instrumental options, in particular the telescope size and sensitivity level, and review the great, transformative potential of the mission as proposed. Specifically, we assess the impact on a broad range of fundamental parameters of our Universe as a function of the expected CMB characteristics, with other papers in the series focusing on controlling astrophysical and instrumental residual systematics. In this paper, we assume that only a few central CORE frequency channels are usable for our purpose, all others being devoted to the cleaning of astrophysical contaminants. On the theoretical side, we assume ΛCDM as our general framework and quantify the improvement provided by CORE over the current constraints from the Planck 2015 release. We also study the joint sensitivity of CORE and of future Baryon Acoustic Oscillation and Large Scale Structure experiments like DESI and Euclid. Specific constraints on the physics of inflation are presented in another paper of the series. In addition to the six parameters of the base ΛCDM, which describe the matter content of a spatially flat universe with adiabatic and scalar primordial fluctuations from inflation, we derive the precision achievable on parameters like those describing curvature, neutrino physics, extra light relics, primordial helium abundance, dark matter annihilation, recombination physics, variation of fundamental constants, dark energy, modified gravity, reionization and cosmic birefringence. In addition to assessing the improvement on the precision of individual parameters, we also forecast the post-CORE overall reduction of the allowed

Reactor physics of CANFLEX

International Nuclear Information System (INIS)

Sim, K. S.; Min, Byung Joo.

1997-07-01

Characteristic of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety. Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. This report also describes about the status of critical assemblies in other countries. (author). 58 refs., 41 tabs., 126 figs

Vver-1000 Mox core computational benchmark

International Nuclear Information System (INIS)

2006-01-01

The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

The Fuzziness of Giant Planets’ Cores

Energy Technology Data Exchange (ETDEWEB)

Helled, Ravit [Institute for Computational Science, University of Zurich, Zurich (Switzerland); Stevenson, David [Division of Geological and Planetary Sciences, Caltech, Pasadena, CA (United States)

2017-05-01

Giant planets are thought to have cores in their deep interiors, and the division into a heavy-element core and hydrogen–helium envelope is applied in both formation and structure models. We show that the primordial internal structure depends on the planetary growth rate, in particular, the ratio of heavy elements accretion to gas accretion. For a wide range of likely conditions, this ratio is in one-to-one correspondence with the resulting post-accretion profile of heavy elements within the planet. This flux ratio depends sensitively on the assumed solid-surface density in the surrounding nebula. We suggest that giant planets’ cores might not be distinct from the envelope and includes some hydrogen and helium, and the deep interior can have a gradual heavy-element structure. Accordingly, Jupiter’s core may not be well defined. Accurate measurements of Jupiter’s gravitational field by Juno could put constraints on Jupiter’s core mass. However, as we suggest here, the definition of Jupiter’s core is complex, and the core’s physical properties (mass, density) depend on the actual definition of the core and on the planet’s growth history.

The Fuzziness of Giant Planets’ Cores

International Nuclear Information System (INIS)

Helled, Ravit; Stevenson, David

2017-01-01

Giant planets are thought to have cores in their deep interiors, and the division into a heavy-element core and hydrogen–helium envelope is applied in both formation and structure models. We show that the primordial internal structure depends on the planetary growth rate, in particular, the ratio of heavy elements accretion to gas accretion. For a wide range of likely conditions, this ratio is in one-to-one correspondence with the resulting post-accretion profile of heavy elements within the planet. This flux ratio depends sensitively on the assumed solid-surface density in the surrounding nebula. We suggest that giant planets’ cores might not be distinct from the envelope and includes some hydrogen and helium, and the deep interior can have a gradual heavy-element structure. Accordingly, Jupiter’s core may not be well defined. Accurate measurements of Jupiter’s gravitational field by Juno could put constraints on Jupiter’s core mass. However, as we suggest here, the definition of Jupiter’s core is complex, and the core’s physical properties (mass, density) depend on the actual definition of the core and on the planet’s growth history.

HYDRATE CORE DRILLING TESTS

Energy Technology Data Exchange (ETDEWEB)

John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

2002-11-01

The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate

Core design and fuel management studies

International Nuclear Information System (INIS)

Min, Byung Joo; Chan, P.

1997-06-01

The design target for the CANDU 9 requires a 20% increase in electrical power output from an existing 480-channel CANDU core. Assuming a net electrical output of 861 MW(e) for a natural uranium fuelled Bruce-B/Darlington reactor in a warm water site, the net electrical output of the reference CANDU 9 reactor would be 1033 MW(e). This report documents the result of the physics studies for the design of the CANDU 9 480/SEU core. The results of the core design and fuel management studies of the CANDU 9 480/SEU reactor indicated that up to 1033 MW(e) output can be achieved in a 480-channel CANDU core by using SEU core can easily be maintained indefinitely using an automated refuelling program. Fuel performance evaluation based on the data of the 500 FPDs refuelling simulation concluded that SEU fuel failure is not expected. (author). 2 tabs., 38 figs., 5 refs

Relative and Absolute Reliability of the Professionalism in Physical Therapy Core Values Self-Assessment Tool.

Science.gov (United States)

Furgal, Karen E; Norris, Elizabeth S; Young, Sonia N; Wallmann, Harvey W

2018-01-01

Development of professional behaviors in Doctor of Physical Therapy (DPT) students is an important part of professional education. The American Physical Therapy Association (APTA) has developed the Professionalism in Physical Therapy Core Values Self-Assessment (PPTCV-SA) tool to increase awareness of personal values in practice. The PPTCV-SA has been used to measure growth in professionalism following a clinical or educational experience. There are few studies reporting psychometric properties of the PPTCV-SA. The purpose of this study was to establish properties of relative reliability (intraclass correlation coefficient, iCC) and absolute reliability (standard error of measurement, SEM; minimal detectable change, MDC) of the PPTCV-SA. in this project, 29 first-year students in a DPT program were administered the PPTCVA-SA on two occasions, 2 weeks apart. Paired t-tests were used to examine stability in PPTCV-SA scores on the two occasions. iCCs were calculated as a measure of relative reliability and for use in the calculation of the absolute reliability measures of SEM and MDC. Results of paired t-tests indicated differences in the subscale scores between times 1 and 2 were non-significant, except for three subscales: Altruism (p=0.01), Excellence (p=0.05), and Social Responsibility (p=0.02). iCCs for test-retest reliability were moderate-to-good for all subscales, with SEMs ranging from 0.30 to 0.62, and MDC95 ranging from 0.83 to 1.71. These results can guide educators and researchers when determining the likelihood of true change in professionalism following a professional development activity.

Opacity in compact extragalactic radio sources and the core shift effect

International Nuclear Information System (INIS)

Kovalev, Y Y; Lobanov, A P; Pushkarev, A B; Zensus, J A

2008-01-01

The apparent position of the 'core' in a parsec-scale radio jet (a compact, bright emitting region at the narrow end of the jet) depends on the observing frequency, owing to synchrotron self-absorption and external absorption. This dependency both provides a tool to probe physical conditions in the vicinity of the core and poses problems for astrometric studies using compact radio sources. We investigate the frequency-dependent shift of the positions of the cores (core shift) observed with very long baseline interferometry (VLBI) in parsec-scale jets. We present results for 29 selected active galactic nuclei (AGN). In these AGN, the magnitude of the measured core shift between 2.3 and 8.6 GHz reaches 1.4 mas, with a median value for the sample of 0.44 mas. We discuss related physics as well as astrometry applications and plans for further studies.

A physical tool for severe accident mitigation studies

Energy Technology Data Exchange (ETDEWEB)

Marie, N., E-mail: nathalie.marie@cea.fr [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Bachrata, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France); Barjot, F. [EDF R& D, SINETICS, F-93141 Clamart (France); Marrel, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Gossé, S. [CEA, DEN, DPC, F-91191 Gif Sur Yvette (France); Bertrand, F. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France)

2016-12-01

Highlights: • Physical tool for mitigation studies devoted to SFR safety. • Physical models to describe the material discharge from core. • Comparison to SIMMER III results. • Studies for ASTRID safety assessment and support to core design. - Abstract: Within the framework of the Generation IV Sodium-cooled Fast Reactors (SFR) R&D program of CEA, the core behavior in case of severe accidents is being assessed. Such transients are usually simulated with mechanistic codes (such as SIMMER-III). As a complement to this code, which gives reference accidental transient, a physico-statistical approach is currently followed; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running simulation of extended accident sequences coupling low-dimensional physical models to advanced statistical analysis techniques. In this context, this paper presents such a low-dimensional physical tool (models and simulation results) dedicated to molten core materials discharge. This 0D tool handles heat transfers from molten (possibly boiling) pools, fuel crust evolution, phase separation/mixing of fuel/steel pools, radial thermal erosion of mitigation tubes, discharge of core materials and associated axial thermal erosion of mitigation tubes. All modules are coupled with a global neutronic evolution model of the degraded core. This physical tool is used to study and to define mitigation features (function of tubes devoted to mitigation inside the core, impact of absorbers falling into the degraded core…) to avoid energetic core recriticality during a secondary phase of a potential severe accident. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties would enable sensitivity analysis studies. This physical tool is described before presenting its comparison against SIMMER-III code results, including a space-and energy-dependent neutron transport kinetic

Start of the construction of initial advanced BWRs

International Nuclear Information System (INIS)

Yamada, Yasuo; Furukawa, Hideyasu; Yoshida, Tomiharu

1992-01-01

The construction of first advanced BWRs (ABWR), No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., was started, and the commercial operation is scheduled to begin in 1996 and 1997. The ABWR with the unit capacity of 1356 MW which is the largest class in the world has been developed aiming at the further improvement of safety, operation performance, economical efficiency and so on. Hitachi Limited carried out the detailed review on new parts design and expected the perfection so as to make the first machine into the plant maintaining high reliability. The course of development of the ABWR, the outline of No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, the review of the design of new parts, the establishment of the technology of manufacturing RCCVs, the verification of the maintenance and checkup of internal pumps and improved type control rod driving mechanism and so on are reported. The ABWR is Japanese third improved type standard plant, but also it is expected to be adopted worldwide. (K.I.)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

In-core fuel management benchmarks for PHWRs

International Nuclear Information System (INIS)

1996-06-01

Under its in-core fuel management activities, the IAEA set up two co-ordinated research programmes (CRPs) on complete in-core fuel management code packages. At a consultant meeting in November 1988 the outline of the CRP on in-core fuel management benchmars for PHWRs was prepared, three benchmarks were specified and the corresponding parameters were defined. At the first research co-ordination meeting in December 1990, seven more benchmarks were specified. The objective of this TECDOC is to provide reference cases for the verification of code packages used for reactor physics and fuel management of PHWRs. 91 refs, figs, tabs

The American Society for Radiation Oncology's 2010 core physics curriculum for radiation oncology residents.

Science.gov (United States)

Xiao, Ying; Bernstein, Karen De Amorim; Chetty, Indrin J; Eifel, Patricia; Hughes, Lesley; Klein, Eric E; McDermott, Patrick; Prisciandaro, Joann; Paliwal, Bhudatt; Price, Robert A; Werner-Wasik, Maria; Palta, Jatinder R

2011-11-15

In 2004, the American Society for Radiation Oncology (ASTRO) published its first physics education curriculum for residents, which was updated in 2007. A committee composed of physicists and physicians from various residency program teaching institutions was reconvened again to update the curriculum in 2009. Members of this committee have associations with ASTRO, the American Association of Physicists in Medicine, the Association of Residents in Radiation Oncology, the American Board of Radiology (ABR), and the American College of Radiology. Members reviewed and updated assigned subjects from the last curriculum. The updated curriculum was carefully reviewed by a representative from the ABR and other physics and clinical experts. The new curriculum resulted in a recommended 56-h course, excluding initial orientation. Learning objectives are provided for each subject area, and a detailed outline of material to be covered is given for each lecture hour. Some recent changes in the curriculum include the addition of Radiation Incidents and Bioterrorism Response Training as a subject and updates that reflect new treatment techniques and modalities in a number of core subjects. The new curriculum was approved by the ASTRO board in April 2010. We anticipate that physicists will use this curriculum for structuring their teaching programs, and subsequently the ABR will adopt this educational program for its written examination. Currently, the American College of Radiology uses the ASTRO curriculum for their training examination topics. In addition to the curriculum, the committee updated suggested references and the glossary. The ASTRO physics education curriculum for radiation oncology residents has been updated. To ensure continued commitment to a current and relevant curriculum, the subject matter will be updated again in 2 years. Copyright © 2011 Elsevier Inc. All rights reserved.

Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

International Nuclear Information System (INIS)

Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

2008-01-01

As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

A passive decay heat removal system for LWRs based on air cooling

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

2015-05-15

Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

Turbine component casting core with high resolution region

Science.gov (United States)

Kamel, Ahmed; Merrill, Gary B.

2014-08-26

A hollow turbine engine component with complex internal features can include a first region and a second, high resolution region. The first region can be defined by a first ceramic core piece formed by any conventional process, such as by injection molding or transfer molding. The second region can be defined by a second ceramic core piece formed separately by a method effective to produce high resolution features, such as tomo lithographic molding. The first core piece and the second core piece can be joined by interlocking engagement that once subjected to an intermediate thermal heat treatment process thermally deform to form a three dimensional interlocking joint between the first and second core pieces by allowing thermal creep to irreversibly interlock the first and second core pieces together such that the joint becomes physically locked together providing joint stability through thermal processing.

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

Energy Technology Data Exchange (ETDEWEB)

Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

1983-08-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

International Nuclear Information System (INIS)

Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

1983-01-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Bringing the Pieces Together – Placing Core Facilities at the Core of Universities and Institutions: Lessons from Mergers, Acquisitions and Consolidations

Science.gov (United States)

Mundoma, Claudius

2013-01-01

As organizations expand and grow, the core facilities have become more dispersed disconnected. This is happening at a time when collaborations within the organization is a driver to increased productivity. Stakeholders are looking at the best way to bring the pieces together. It is inevitable that core facilities at universities and research institutes have to be integrated in order to streamline services and facilitate ease of collaboration. The path to integration often goes through consolidation, merging and shedding of redundant services. Managing this process requires a delicate coordination of two critical factors: the human (lab managers) factor and the physical assets factor. Traditionally more emphasis has been placed on reorganizing the physical assets without paying enough attention to the professionals who have been managing the assets for years, if not decades. The presentation focuses on how a systems approach can be used to effect a smooth core facility integration process. Managing the human element requires strengthening existing channels of communication and if necessary, creating new ones throughout the organization to break cultural and structural barriers. Managing the physical assets requires a complete asset audit and this requires direct input from the administration as well as the facility managers. Organizations can harness the power of IT to create asset visibility. Successfully managing the physical assets and the human assets increases productivity and efficiency within the organization.

Inner core tilt and polar motion

Science.gov (United States)

Dumberry, Mathieu; Bloxham, Jeremy

2002-11-01

A tilted inner core permits exchange of angular momentum between the core and the mantle through gravitational and pressure torques and, as a result, changes in the direction of Earth's axis of rotation with respect to the mantle. We have developed a model to calculate the amplitude of the polar motion that results from an equatorial torque at the inner core boundary which tilts the inner core out of alignment with the mantle. We specifically address the issue of the role of the inner core tilt in the decade polar motion known as the Markowitz wobble. We show that a decade polar motion of the same amplitude as the observed Markowitz wobble requires a torque of 1020 N m which tilts the inner core by 0.07 degrees. This result critically depends on the viscosity of the inner core; for a viscosity less than 5 × 1017 Pa s, larger torques are required. We investigate the possibility that a torque of 1020 N m with decadal periodicity can be produced by electromagnetic coupling between the inner core and torsional oscillations of the flow in the outer core. We demonstrate that a radial magnetic field at the inner core boundary of 3 to 4 mT is required to obtain a torque of such amplitude. The resulting polar motion is eccentric and polarized, in agreement with the observations. Our model suggests that equatorial torques at the inner core boundary might also excite the Chandler wobble, provided there exists a physical mechanism that can generate a large torque at a 14 month period.

Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

OpenAIRE

Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

2010-01-01

Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

Characteristic test of initial HTTR core

International Nuclear Information System (INIS)

Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

2004-01-01

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

A programmatic approach for implementing MOX fuel operation in advanced and existing boiling water reactors

International Nuclear Information System (INIS)

Ehrlich, E.H.; Knecht, P.D.; Shirley, N.C.; Wadekamper, D.C.

1996-01-01

This paper describes a programmatic overview of the elements and issues associated with MOX fuel utilization. Many of the dominant considerations and integrated relationships inherent in initiating MOX fuel utilization in BWRs or the ABWR with partial or full MOX core designs are discussed. The most significant considerations in carrying out a MOX implementation program, while achieving commercially desirable fuel cycles and commercially manageable MOX fuel fabrication, testing, qualification, and licensing support activities, are described. The impact of politics and public influences and the necessary role of industry and government contributions are also discussed. (J.P.N.)

Efficiency of different techniques of physical flattening by fuel while selection of optimum arrangement of large fast reactor core

International Nuclear Information System (INIS)

Grachev, E.A.; Dejnega, N.L.; Mitin, A.M.

1974-01-01

Results are given of calculations for selecting the parameters of the large fast breeder reactor core (1500 Mw) operating on oxide fuel with a sodium coolant. A complex optimum criterion was selected for energy intensity, energy distribution, breeding ratio and critical load factor, run duration, burning, reactivity variations, influence of CV3, fuel overloads, and calculated fue fuel expenses. The effectivities of various methods for physical grading of fuel (enrichment and composition) were examined in accordance with the optimum criterion. Parameters of reactor cores optimum arrangements are presented. Continuous reactor operation during 0.8-1.0 yr. at energy intensity more than 400 kW was shown to be essential for attaining the optimum chosen. Accounting for the CV3 system and partial fuel overloads, the methods of balancing energy release either by enriching fuel or by changing its composition proved to be almost equally effective. All calculations were performed with the aid of a 18-4-RZ-15B program on the basis of a BNAB-26 constant system [ru

Understanding physics

CERN Document Server

Mansfield, Michael

2011-01-01

Understanding Physics - Second edition is a comprehensive, yet compact, introductory physics textbook aimed at physics undergraduates and also at engineers and other scientists taking a general physics course. Written with today's students in mind, this text covers the core material required by an introductory course in a clear and refreshing way. A second colour is used throughout to enhance learning and understanding. Each topic is introduced from first principles so that the text is suitable for students without a prior background in physics. At the same time the book is designed to enable

Activity Report of Reactor Physics Division - 1997

International Nuclear Information System (INIS)

Singh, Om Pal

1998-01-01

The research and development activities of the Reactor Physics Division of the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1997 are reported. The activities are arranged under the headings: nuclear data processing and validation, PFBR and KAMINI core physics, FBTR core physics, radioactivity and shielding and safety analysis. A list of publications of the Division and seminars delivered are included at the end of the report

Lead users' ideas on core features to support physical activity in rheumatoid arthritis: a first step in the development of an internet service using participatory design.

Science.gov (United States)

Revenäs, Åsa; Opava, Christina H; Åsenlöf, Pernilla

2014-03-22

Despite the growing evidence of the benefits of physical activity (PA) in individuals with rheumatoid arthritis (RA), the majority is not physically active enough. An innovative strategy is to engage lead users in the development of PA interventions provided over the internet. The aim was to explore lead users' ideas and prioritization of core features in a future internet service targeting adoption and maintenance of healthy PA in people with RA. Six focus group interviews were performed with a purposively selected sample of 26 individuals with RA. Data were analyzed with qualitative content analysis and quantification of participants' prioritization of most important content. Six categories were identified as core features for a future internet service: up-to-date and evidence-based information and instructions, self-regulation tools, social interaction, personalized set-up, attractive design and content, and access to the internet service. The categories represented four themes, or core aspects, important to consider in the design of the future service: (1) content, (2) customized options, (3) user interface and (4) access and implementation. This is, to the best of our knowledge, the first study involving people with RA in the development of an internet service to support the adoption and maintenance of PA.Participants helped identifying core features and aspects important to consider and further explore during the next phase of development. We hypothesize that involvement of lead users will make transfer from theory to service more adequate and user-friendly and therefore will be an effective mean to facilitate PA behavior change.

PAW [Physics Analysis Workstation] at Fermilab: CORE based graphics implementation of HIGZ [High Level Interface to Graphics and Zebra

International Nuclear Information System (INIS)

Johnstad, H.

1989-06-01

The Physics Analysis Workstation system (PAW) is primarily intended to be the last link in the analysis chain of experimental data. The graphical part of PAW is based on HIGZ (High Level Interface to Graphics and Zebra), which is based on the OSI and ANSI standard Graphics Kernel System (GKS). HIGZ is written in the context of PAW. At Fermilab, the CORE based graphics system DI-3000 by Precision Visuals Inc., is widely used in the analysis of experimental data. The graphical part of the PAW routines has been totally rewritten and implemented in the Fermilab environment. 3 refs

The MCART radiation physics core: the quest for radiation dosimetry standardization.

Science.gov (United States)

Kazi, Abdul M; MacVittie, Thomas J; Lasio, Giovanni; Lu, Wei; Prado, Karl L

2014-01-01

Dose-related radiobiological research results can only be compared meaningfully when radiation dosimetry is standardized. To this purpose, the National Institute of Allergy and Infectious Diseases (NIAID)-sponsored Medical Countermeasures Against Radiological Threats (MCART) consortium recently created a Radiation Physics Core (RPC) as an entity to assume responsibility of standardizing radiation dosimetry practices among its member laboratories. The animal research activities in these laboratories use a variety of ionizing photon beams from several irradiators such as 250-320 kVp x-ray generators, Cs irradiators, Co teletherapy machines, and medical linear accelerators (LINACs). In addition to this variety of sources, these centers use a range of irradiation techniques and make use of different dose calculation schemes to conduct their experiments. An extremely important objective in these research activities is to obtain a Dose Response Relationship (DRR) appropriate to their respective organ-specific models of acute and delayed radiation effects. A clear and unambiguous definition of the DRR is essential for the development of medical countermeasures. It is imperative that these DRRs are transparent between centers. The MCART RPC has initiated the establishment of standard dosimetry practices among member centers and is introducing a Remote Dosimetry Monitoring Service (RDMS) to ascertain ongoing quality assurance. This paper will describe the initial activities of the MCART RPC toward implementing these standardization goals. It is appropriate to report a summary of initial activities with the intent of reporting the full implementation at a later date.

GNPS 18-months fuel cycles core thermal hydraulic design

International Nuclear Information System (INIS)

Liu Changwen; Zhou Zhou

2002-01-01

GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

Current status of water chemistry in Japan

Energy Technology Data Exchange (ETDEWEB)

Ishigure, K. [Saitama Inst. of Tech. (Japan)

2002-07-01

At present 28 BWRs including 2 ABWRs and 23 PWRs are in operation in Japan and generated 36.8{open_square} of total electric power in 1998. Totally 4 BWRs, of which two are ABWRs, are now under construction, and one BWR together with one ABWR is in the stage of planning. One gas-cooled reactor (Tokai-1) was shut down permanently in 1998 and last year entered into decommissioning stage. According to the Japanese 2001 plan of electric power supply, 13 nuclear power plants newly constructed are to start operation in the next 10 years. In this paper the recent status of water chemistry technology in Japanese nuclear power plants is briefly summarized together with a touch upon the activities in the fundamental research. (author)

Current status of water chemistry in Japan

International Nuclear Information System (INIS)

Ishigure, K.

2002-01-01

At present 28 BWRs including 2 ABWRs and 23 PWRs are in operation in Japan and generated 36.8□ of total electric power in 1998. Totally 4 BWRs, of which two are ABWRs, are now under construction, and one BWR together with one ABWR is in the stage of planning. One gas-cooled reactor (Tokai-1) was shut down permanently in 1998 and last year entered into decommissioning stage. According to the Japanese 2001 plan of electric power supply, 13 nuclear power plants newly constructed are to start operation in the next 10 years. In this paper the recent status of water chemistry technology in Japanese nuclear power plants is briefly summarized together with a touch upon the activities in the fundamental research. (author)

Software for physical start-up console

International Nuclear Information System (INIS)

Arbet, L.; Suchy, R.

1991-01-01

The physical start-up console comprises an PC AT-based control unit equipped with an 80386 processor, and information input/output units. The basic functions to be fulfilled by the control unit software include data acquisition related to the following parameters: neutron physics properties of the reactor core (neutron fluxes recorded by ionization chambers and reactivity recorded by a digital reactimeter), positions of the reactor core control elements (by the digital position meter) and reactor core control measurements, and technological quantities requisite for evaluating physical start-up tests. The measured and calculated data are shown on the control unit display. The setup of the data acquisition system and of user programs is dealt with, and characteristics of the user processes are briefly described. (Z.S.)

Analysis of core physics and thermal-hydraulics results of control rod withdrawal experiments in the LOFT facility

International Nuclear Information System (INIS)

Varacalle, D.J. Jr.; Chen, T.H.; Harvego, E.A.; Ollikkala, H.

1983-01-01

Two anticipated transient experiments simulating an uncontrolled control rod withdrawal event in a pressurized water reactor (PWR) were conducted in the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The scaled LOFT 50-MW(t) PWR includes most of the principal features of larger commercial PWRs. The experiments tested the ability of reactor analysis codes to accurately calculate core reactor physics and thermal-hydraulic phenomena in an integral reactor system. The initial conditions and scaled operating parameters for the experiments were representative of those expected in a commercial PWR. In both experiments, all four LOFT control rod assemblies were withdrawn at a reactor power of 37.5 MW and a system pressure of 14.8 MPa

A simple model for induction core voltage distributions

International Nuclear Information System (INIS)

Briggs, Richard J.; Fawley, William M.

2004-01-01

In fall 2003 T. Hughes of MRC used a full EM simulation code (LSP) to show that the electric field stress distribution near the outer radius of the longitudinal gaps between the four Metglas induction cores is very nonuniform in the original design of the DARHT-2 accelerator cells. In this note we derive a simple model of the electric field distribution in the induction core region to provide physical insights into this result. The starting point in formulating our model is to recognize that the electromagnetic fields in the induction core region of the DARHT-2 accelerator cells should be accurately represented within a quasi-static approximation because the timescale for the fields to change is much longer than the EM wave propagation time. The difficulty one faces is the fact that the electric field is a mixture of both a ''quasi-magnetostatic field'' (having a nonzero curl, with Bdot the source) and a ''quasi-electrostatic field'' (the source being electric charges on the various metal surfaces). We first discuss the EM field structure on the ''micro-scale'' of individual tape windings in Section 2. The insights from that discussion are then used to formulate a ''macroscopic'' description of the fields inside an ''equivalent homogeneous tape wound core region'' in Section 3. This formulation explicitly separates the nonlinear core magnetics from the quasi-electrostatic components of the electric field. In Section 4 a physical interpretation of the radial dependence of the electrostatic component of the electric field derived from this model is presented in terms of distributed capacitances, and the voltage distribution from gap to gap is related to various ''equivalent'' lumped capacitances. Analytic solutions of several simple multi-core cases are presented in Sections 5 and 6 to help provide physical insight into the effect of various proposed changes in the geometrical parameters of the DARHT-2 accelerator cell. Our results show that over most of the gap

Reactor core simulations in Canada

International Nuclear Information System (INIS)

Roy, R.; Koclas, J.; Shen, W.; Jenkins, D. A.; Altiparmakov, D.; Rouben, B.

2004-01-01

This review will address the current simulation flow-chart currently used for reactor-physics simulations in the Canadian industry. The neutron behaviour in heavy-water moderated power reactors is quite different from that in other power reactors, thus the core physics approximations are somewhat different Some codes used are particular to the context of heavy-water reactors, and the paper focuses on this aspect. The paper also shows simulations involving new design features of the Advanced Candu Reactor TM (ACR TM), and provides insight into future development, expected in the coming years. (authors)

Chemical and physical changes during seawater flow through intact dunite cores: An experimental study at 150-200 °C

Science.gov (United States)

Luhmann, Andrew J.; Tutolo, Benjamin M.; Bagley, Brian C.; Mildner, David F. R.; Scheuermann, Peter P.; Feinberg, Joshua M.; Ignatyev, Konstantin; Seyfried, William E.

2017-10-01

values for the two pristine samples and the 150 °C post-experimental core that underwent negligible reaction, indicating that any change from reaction was smaller than the natural variability of the dunite. Even though there was little physical evidence of alteration, the initial stage of serpentinization at 200 °C was sufficiently significant to have a dramatic effect on flow fields in the core. Furthermore, this experiment generated significant dissolved hydrogen concentrations while simulating open system dynamics. Even though open systems prevent elevated hydrogen concentrations due to continual loss of hydrogen, we speculate that this process is responsible for stabilizing ferric Fe-rich serpentine in nature while also oxidizing more ferrous iron (Fe(II)) and cumulatively generating more hydrogen than would be possible in a closed system.

Reactor physics and thermodynamics of a gaseous core fission reactor

International Nuclear Information System (INIS)

Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

1990-01-01

Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

User interface design and system integration aspects of core monitoring systems

International Nuclear Information System (INIS)

Berg, O.; Bodal, T.; Hornaes, A.; Porsmyr, J.

2000-01-01

The present paper describes our experience with the SCORPIO core monitoring system using generic building blocks for the MMI and system integration. In this context the different layers of the software system are discussed starting with the communication system, interfacing of various modules (e.g. physics codes), administration of several modules and generation of graphical user interfaces for different categories of end-users. A method by which re-use of software components can make the system development and maintenance more efficient is described. Examples are given from different system installation projects. The methodology adopted is considered particularly important in the future, as it is anticipated that core monitoring systems will be expanded with new functions (e.g. information from technical specifications, procedures, noise analysis, etc). Further, efficient coupling of off-line tools for core physics calculations and on-line modules in core monitoring can pave the way for cost savings. (authors)

HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

International Nuclear Information System (INIS)

Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

2002-01-01

This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

International Nuclear Information System (INIS)

Winter, Dominik

2014-01-01

increase of the fission reaction rate of plutonium in the upper part of the active height leading to an increase of the neutron flux and subsequent induced fission reactions - ThPu: an increase of the fissile inventory ratio with ongoing operation due to the production of the fissile uranium-233 from the fertile thorium-232 Finally, for both core loadings, equal or even more favorable reactor physical safety parameters are achieved which remain within the licensed safety margins of today's BWRs.

Analysis and research status of severe core damage accidents

International Nuclear Information System (INIS)

1984-03-01

The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

The American Society for Radiation Oncology’s 2010 Core Physics Curriculum for Radiation Oncology Residents

International Nuclear Information System (INIS)

Xiao Ying; De Amorim Bernstein, Karen; Chetty, Indrin J.; Eifel, Patricia; Hughes, Lesley; Klein, Eric E.; McDermott, Patrick; Prisciandaro, Joann; Paliwal, Bhudatt; Price, Robert A.; Werner-Wasik, Maria; Palta, Jatinder R.

2011-01-01

Purpose: In 2004, the American Society for Radiation Oncology (ASTRO) published its first physics education curriculum for residents, which was updated in 2007. A committee composed of physicists and physicians from various residency program teaching institutions was reconvened again to update the curriculum in 2009. Methods and Materials: Members of this committee have associations with ASTRO, the American Association of Physicists in Medicine, the Association of Residents in Radiation Oncology, the American Board of Radiology (ABR), and the American College of Radiology. Members reviewed and updated assigned subjects from the last curriculum. The updated curriculum was carefully reviewed by a representative from the ABR and other physics and clinical experts. Results: The new curriculum resulted in a recommended 56-h course, excluding initial orientation. Learning objectives are provided for each subject area, and a detailed outline of material to be covered is given for each lecture hour. Some recent changes in the curriculum include the addition of Radiation Incidents and Bioterrorism Response Training as a subject and updates that reflect new treatment techniques and modalities in a number of core subjects. The new curriculum was approved by the ASTRO board in April 2010. We anticipate that physicists will use this curriculum for structuring their teaching programs, and subsequently the ABR will adopt this educational program for its written examination. Currently, the American College of Radiology uses the ASTRO curriculum for their training examination topics. In addition to the curriculum, the committee updated suggested references and the glossary. Conclusions: The ASTRO physics education curriculum for radiation oncology residents has been updated. To ensure continued commitment to a current and relevant curriculum, the subject matter will be updated again in 2 years.

Methods optimization for the first time core critical

International Nuclear Information System (INIS)

Yan Liang

2014-01-01

The PWR reactor core commissioning programs the content of the first critical reactor physics experiment, and describes thc physical test method. However, all the methods arc not exactly the same but efficient. This article aims to enhance the reactor for the first time in the process of critical safety, shorten the overall time of critical physical test for the first time, and improve the integrity of critical physical test data for the first time and accuracy, eventually to improve the operation of the plant economic benefit adopting sectional dilution, power feedback for Doppler point improvement of physical test methods, and so on. (author)

Effect of Core Training Program on Physical Functional Performance in Female Soccer Players

Science.gov (United States)

Taskin, Cengiz

2016-01-01

The purpose of this study was to determine the effect of core training program on speed, acceleration, vertical jump, and standing long jump in female soccer players. A total of 40 female soccer players volunteered to participate in this study. They were divided randomly into 1 of 2 groups: core training group (CTG; n = 20) and control group (CG;…

Geodynamo Modeling of Core-Mantle Interactions

Science.gov (United States)

Kuang, Wei-Jia; Chao, Benjamin F.; Smith, David E. (Technical Monitor)

2001-01-01

Angular momentum exchange between the Earth's mantle and core influences the Earth's rotation on time scales of decades and longer, in particular in the length of day (LOD) which have been measured with progressively increasing accuracy for the last two centuries. There are four possible coupling mechanisms for transferring the axial angular momentum across the core-mantle boundary (CMB): viscous, magnetic, topography, and gravitational torques. Here we use our scalable, modularized, fully dynamic geodynamo model for the core to assess the importance of these torques. This numerical model, as an extension of the Kuang-Bloxham model that has successfully simulated the generation of the Earth's magnetic field, is used to obtain numerical results in various physical conditions in terms of specific parameterization consistent with the dynamical processes in the fluid outer core. The results show that depending on the electrical conductivity of the lower mantle and the amplitude of the boundary topography at CMB, both magnetic and topographic couplings can contribute significantly to the angular momentum exchange. This implies that the core-mantle interactions are far more complex than has been assumed and that there is unlikely a single dominant coupling mechanism for the observed decadal LOD variation.

Xenon oscillation tests in four-loop PWR cores

International Nuclear Information System (INIS)

Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

1980-01-01

The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

Transmission in Optically Transparent Core Networks

Science.gov (United States)

Kilper, Dan; Jensen, Rich; Petermann, Klaus; Karasek, Miroslav

2007-03-01

Call for Papers: Transmission in Optically Transparent Core Networks Guest Feature Editors Dan Kilper and Rich Jensen, Coordinating Associate Editors Klaus Petermann and Miroslav Karasek, Guest Feature Editors Submission deadline: 15 June 2007 Optically transparent networks in which optical transport signals are routed uninterrupted through multiple nodes have long been viewed as an important evolutionary step in fiber optic communications. More than a decade of research and development on transparent network technologies together with the requisite traffic growth has culminated in the recent deployment of commercial optically transparent systems. Although many of the traditional research goals of optical transmission remain important, optical transparency introduces new challenges. Greater emphasis is placed on system efficiency and control. The goal of minimizing signal terminations, which has been pursued through increasing reach and channel capacity, also can be realized through wavelength routing techniques. Rather than bounding system operation by rigid engineering rules, the physical layer is controlled and managed by automation tools. Many static signal impairments become dynamic due to network reconfiguration and transient fault events. Recently new directions in transmission research have emerged to address transparent networking problems. This special issue of the Journal of Optical Networking will examine the technologies and theory underpinning transmission in optically transparent core networks, including both metropolitan and long haul systems. Scope of Submission The special issue editors are soliciting high-quality original research papers related to transmission in optically transparent core networks. Although this does not include edge networks such as access or enterprise networks, core networks that have access capabilities will be considered in scope as will topics related to the interworking between core and edge networks. The core network

Stellar core collapse and supernova

International Nuclear Information System (INIS)

Wilson, J.R.; Mayle, R.; Woosley, S.E.; Weaver, T.

1985-04-01

Massive stars that end their stable evolution as their iron cores collapse to a neutron star or black hole long been considered good candidates for producing Type II supernovae. For many years the outward propagation of the shock wave produced by the bounce of these iron cores has been studied as a possible mechanism for the explosion. For the most part, the results of these studies have not been particularly encouraging, except, perhaps, in the case of very low mass iron cores or very soft nuclear equations of state. The shock stalls, overwhelmed by photodisintegration and neutrino losses, and the star does not explode. More recently, slow late time heating of the envelope of the incipient neutron star has been found to be capable of rejuvenating the stalled shock and producing an explosion after all. The present paper discusses this late time heating and presents results from numerical calculations of the evolution, core collapse, and subsequent explosion of a number of recent stellar models. For the first time they all, except perhaps the most massive, explode with reasonable choices of input physics. 39 refs., 17 figs., 1 tab

Gas core nuclear thermal rocket engine research and development in the former USSR

International Nuclear Information System (INIS)

Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

1992-09-01

Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

Preliminary analysis of the proposed BN-600 benchmark core

International Nuclear Information System (INIS)

John, T.M.

2000-01-01

The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

Design of radiation shields in nuclear reactor core

International Nuclear Information System (INIS)

Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

2008-01-01

This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

Fuel management study on quarter core refueling for Ling Ao NPP

International Nuclear Information System (INIS)

Zhang Hong; Li Jinggang

2012-01-01

The fuel management study on quarter core refueling is introduced for Ling Ao NPP. Starting from the selection of the objective of fuel management for quarter core refueling, the code and method used and the analysis carried out are explained in details to reach the final loading pattern chosen. The start-up physics test results are listed to demonstrate the realized quarter core fuel management. In the end, the advantage and disadvantage after turning to quarter core refueling has been given for the power plant from the fuel management point of view. (authors)

Gravitational wave generation by stellar core collapse

International Nuclear Information System (INIS)

Moore, T.A.

1981-01-01

Stars which have masses greater than 5 to 8 solar masses are thought to undergo a stage of catastrophic core collapse and subsequent supernova explosion at the end of their lives. If the core is not spherically symmetric, the bounce which halts its collapse at transnuclear densities will generate a pulse of gravitational waves. This thesis presents a fully relativistic model of core collapse which treats deviations from spherical symmetry as small perturbations on a spherical background. This model may be used to predict qualitative and quantitative features of the gravitational radiation emitted by stellar cores with odd-parity, axisymmetric fluid perturbations, and represents a first step in the application of perturbative methods to more general asymmetries. The first chapter reviews the present consensus on the physics of core collapse and outlines the important features, assumptions, and limitations of the model. A series of model runs are presented and discussed. Finally, several proposals for future research are presented. Subsequent chapters explore in detail the mathematical features of the present model and its realization on the computer

Preliminary core design of IRIS-50

International Nuclear Information System (INIS)

Petrovic, Bojan; Franceschini, Fausto

2009-01-01

IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

International Nuclear Information System (INIS)

Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

1980-01-01

Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, Jozef; Vocka, Radim

2010-01-01

The SCORPIO (SCORPIO-VVER) core monitoring system, its basic features and history of implementation at Czech NPPs are described. The most important improvements in the area of neutron physics, core thermal analysis and operation support are as follows: Moving to the 42 axial nodes across the whole system (2004); Implementation of new cross section library to support mixed reactor core with differences in axial geometry of used fuel types and enhancement of Core Simulator boundary conditions model, to properly address the 'wild' geometry in axial direction; Adjusting the thermohydraulic and neutron-physical models regarding to the Gd2 fuel needs; Support up to 5 types of FAs and 2 types of SPND (Posit, IST); Extension of form functions for pin-wise reconstruction to improve pin-power prediction in control rod coupler region; System adaptation to the new upgraded digital I and C unit system; Integration of the SCORPIO-VVER system and its workstation into the plant redundant in-core system; Implementation of new On-Line form function generation to module RECON; New design of the Strategy Generator with advanced predictions; Adaptation of the system to support the new up-rated reactor thermal power; Adding new online SDM calculation function into to system; Implementation of the new 3D power reconstruction with SPND interpretation; Extending the limit checking to the 'full core' checking. The control of margins to the technical specification: Extended to full core - all FA is controlled individually in core; The limits are definable up to 59 FA (1/6 symmetry); 4 limited parameters are controlled - Kr, qlin, Tout-fa, dTfa; 2 additional parameters are monitored - dTsat, DNBR; New MMI are developed to present the limited and controlled parameters in core. Upgrade 3 is planned for the Slovak Bohunice NPP in 2011-2012. (P.A.)

Theoretical physics 8 statistical physics

CERN Document Server

Nolting, Wolfgang

2018-01-01

This textbook offers a clear and comprehensive introduction to statistical physics, one of the core components of advanced undergraduate physics courses. It follows on naturally from the previous volumes in this series, using methods of probability theory and statistics to solve physical problems. The first part of the book gives a detailed overview on classical statistical physics and introduces all mathematical tools needed. The second part of the book covers topics related to quantized states, gives a thorough introduction to quantum statistics, followed by a concise treatment of quantum gases. Ideally suited to undergraduate students with some grounding in quantum mechanics, the book is enhanced throughout with learning features such as boxed inserts and chapter summaries, with key mathematical derivations highlighted to aid understanding. The text is supported by numerous worked examples and end of chapter problem sets. About the Theoretical Physics series Translated from the renowned and highly successf...

A volatile-rich Earth's core inferred from melting temperature of core materials

Science.gov (United States)

Morard, G.; Andrault, D.; Antonangeli, D.; Nakajima, Y.; Auzende, A. L.; Boulard, E.; Clark, A. N.; Lord, O. T.; Cervera, S.; Siebert, J.; Garbarino, G.; Svitlyk, V.; Mezouar, M.

2016-12-01

Planetary cores are mainly constituted of iron and nickel, alloyed with lighter elements (Si, O, C, S or H). Understanding how these elements affect the physical and chemical properties of solid and liquid iron provides stringent constraints on the composition of the Earth's core. In particular, melting curves of iron alloys are key parameter to establish the temperature profile in the Earth's core, and to asses the potential occurrence of partial melting at the Core-Mantle Boundary. Core formation models based on metal-silicate equilibration suggest that Si and O are the major light element components1-4, while the abundance of other elements such as S, C and H is constrained by arguments based on their volatility during planetary accretion5,6. Each compositional model implies a specific thermal state for the core, due to the different effect that light elements have on the melting behaviour of Fe. We recently measured melting temperatures in Fe-C and Fe-O systems at high pressures, which complete the data sets available both for pure Fe7 and other binary alloys8. Compositional models with an O- and Si-rich outer core are suggested to be compatible with seismological constraints on density and sound velocity9. However, their crystallization temperatures of 3650-4050 K at the CMB pressure of 136 GPa are very close to, if not higher than the melting temperature of the silicate mantle and yet mantle melting above the CMB is not a ubiquitous feature. This observation requires significant amounts of volatile elements (S, C or H) in the outer core to further reduce the crystallisation temperature of the core alloy below that of the lower mantle. References 1. Wood, B. J., et al Nature 441, 825-833 (2006). 2. Siebert, J., et al Science 339, 1194-7 (2013). 3. Corgne, A., et al Earth Planet. Sc. Lett. 288, 108-114 (2009). 4. Fischer, R. a. et al. Geochim. Cosmochim. Acta 167, 177-194 (2015). 5. Dreibus, G. & Palme, H. Geochim. Cosmochim. Acta 60, 1125-1130 (1995). 6. Mc

Reference Monte Carlo calculations of Maria reactor core

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

2002-01-01

The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

Physics at FAIR

International Nuclear Information System (INIS)

Chattopadhyay, Subhasis

2014-01-01

The Facility for Antiproton and Ion Research (FAIR) is under construction at Darmstadt, Germany. It will deliver high intensity beams of ions and antiprotons for experiments in the fields of atomic physics, plasma physics, nuclear physics, hadron physics, nuclear matter physics, material physics and biophysics. One of the scientific pillars of FAIR is the Compressed Baryonic Matter (CBM) experiment which is designed for the study of high density nuclear matter as it exists in the core of neutron stars. In this article the scientific program of FAIR will be reviewed with emphasis on the CBM experiment

Status of degraded core issues. Synthesis paper prepared by G. Bandini in collaboration with the NEA task group on degraded core cooling

International Nuclear Information System (INIS)

2001-02-01

The in-vessel evolution of a severe accident in a nuclear reactor is characterised, generally, by core uncover and heat-up, core material oxidation and melting, molten material relocation and debris behaviour in the lower plenum up to vessel failure. The in-vessel core melt progression involves a large number of physical and chemical phenomena that may depend on the severe accident sequence and the reactor type under consideration. Core melt progression has been studied in the last twenty years through many experimental works. Since then, computer codes are being developed and validated to analyse different reactor accident sequences. The experience gained from the TMI-2 accident also constitutes an important source of data. The understanding of core degradation process is necessary to evaluate initial conditions for subsequent phases of the accident (ex-vessel and within the containment), and define accident management strategies and mitigative actions for operating and advanced reactors. This synthesis paper, prepared within the Task Group on Degraded Core Cooling (TG-DCC) of PWG2, contains a brief summary of current views on the status of degraded core issues regarding light water reactors. The in-vessel fission product release and transport issue is not addressed in this paper. The areas with remaining uncertainties and the needs for further experimental investigation and model development have been identified. The early phase of core melt progression is reasonably well understood. Remaining uncertainties may be addressed on the basis of ongoing experimental activities, e.g. on core quenching, and research programs foreseen in the near future. The late phase of core melt progression is less understood. Ongoing research programs are providing additional valuable information on corium molten pool behaviour. Confirmatory research is still required. The pool crust behaviour and material relocation into the lower plenum are the areas where additional research should

Rock Physics

DEFF Research Database (Denmark)

Fabricius, Ida Lykke

2017-01-01

Rock physics is the discipline linking petrophysical properties as derived from borehole data to surface based geophysical exploration data. It can involve interpretation of both elastic wave propagation and electrical conductivity, but in this chapter focus is on elasticity. Rock physics is based...... on continuum mechanics, and the theory of elasticity developed for statics becomes the key to petrophysical interpretation of velocity of elastic waves. In practice, rock physics involves interpretation of well logs including vertical seismic profiling (VSP) and analysis of core samples. The results...

Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

Science.gov (United States)

Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

2010-10-01

The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

Influence of FA pin power minimisation on neutron-physical core characteristics

International Nuclear Information System (INIS)

Mikolas, P.

2005-01-01

The influence of optimisation of radial fuel enrichment profilation in fuel assembly of Gd-2 type on lowering its pin power non-uniformity and subsequent lowering of F dH in WWER-440 cores are described in this paper (Author)

Physical events that occur in the reactor core during load changes; Les effets physiques sur le coeur mis en jeu lors des variations de puissance

Energy Technology Data Exchange (ETDEWEB)

Paulin, Ph. [Electricite de France (EDF/DPN/UNIE/GECC), 93 - Saint-Denis (France); Golfier, H. [CEA Saclay (DEN-DANS/DM2S/SERMA/LPEC), 91 - Gif-sur-Yvette (France)

2007-05-15

The reactor core control aims at mastering 2 important parameters that are relevant for reactor availability and safety. First, the reactivity that sets the power output and secondly, the power map in order to handle hot spots. In PWR-type reactors, physical events such as moderator or fuel temperature changes, xenon concentration, that are important for both parameters, evolve during load changes but also during power plateaus and are dependent on burn-up. In this article temperature effect and xenon poisoning are analysed and their impact are assessed along an irradiation campaign through a core neutronic simulation and data from instrumentation. Xenon oscillations are particularly well illustrated. The counter-reactions of the means used for reactor controlling: soluble boron and control rods, are also analysed. (A.C.)

Physics and behaviour during a ULOF of an innovative heterogeneous annular FBR core

International Nuclear Information System (INIS)

Massara, S.; Verwaerde, D.

2012-01-01

The major conclusions: • The reduction of the Na void worth is a way allowing a strong improvement of the dynamic behavior in very severe ULOF transient (10 s halving time), possibly allowing to avoid Na boiling; • 1st order effects: Na density ( 0); • 2nd order effects: - Mass flow gaggling scheme (as a function of the core neutronics); - Other feed-back effects: diagrid, driveline feed-back. → Very strong impact of uncertainties: Thermalhydraulic models & codes, drive-line feed-back modeling; • Methodology for feed-back coefficient calculation (example: in this calculation the Na density effect is linearized from nominal to 100% void, anticonservative in case of no Na boiling); • Core neutronics: nuclear data, models. → Even in case of no Na boiling, the critical events will be: • Fuel cladding and S/A wrapper behavior at very high temperature; • Upper core structures behavior

Radiation exposure reduction technologies for a Japanese Advanced BWR (Dose Rate Reduction Experience in Shika Unit 2)

International Nuclear Information System (INIS)

Saito, Takeshi; Ichikawa, Koji; Ishimaru, Hiroshi; Aizawa, Motohiro; Sato, Yoshiteru; Morita, Shoichi

2012-09-01

Operating experiences of the advanced boiling water reactor (ABWR) have been accumulated in Japan since the first ABWRs Kashiwazaki-Kariwa NPS unit 6 and 7 came into service. Shika NPS unit 2 (Shika-2) of Hokuriku Electric Power Co. is the fourth ABWR plant in Japan. Since ABWRs have no piping of the reactor recirculation system (RRS), which is the largest source of radiation in conventional BWRs, carbon steel piping of the reactor water cleanup system (RWCU) and residual heat removal system (RHR) are the largest source in ABWRs. Therefore we have focused on reduction methods of radioactive material on carbon steel surface in order to reduce the quantity of occupational exposure in Shika-2. In Shika-2 the following methods have been adopted. Before fuel loading, alkaline pre-filming process was applied to the RWCU piping during plant startup testing. After start of operation, the feed water iron concentration control (or nickel/iron ratio control) method was applied. Furthermore, during shutdown operation the RHR system was operated when the reactor water temperature was dropped to 120 degree C with the use of condenser heat sink cooling operation. These dose rate reduction methods worked well in Shika-2 as expected. The quantity of occupational exposure at the 3 rd outage in Shika-2 was approximately 0.35 person-Sv. On the basis of the results obtained thus far, the occupational dose expected at the outage after deposition amount of radioactivity reaching the equilibrium state has been estimated to be around 0.5 person-Sv. This value is considered to be in low level compared with the worldwide statistics. (authors)

First results from core-edge parallel composition in the FACETS project.

Energy Technology Data Exchange (ETDEWEB)

Cary, J. R.; Candy, J.; Cohen, R. H.; Krasheninnikov, S.; McCune, D. C.; Estep, D. J.; Larson, J.; Malony, A. D.; Pankin, A.; Worley, P. H.; Carlsson, J. A.; Hakim, A. H.; Hamill, P.; Kruger, S.; Miah, M.; Muzsala, S.; Pletzer, A.; Shasharina, S.; Wade-Stein, D.; Wang, N.; Balay, S.; McInnes, L.; Zhang, H.; Casper, T.; Diachin, L. (Mathematics and Computer Science); (Tech-X Corp.); (General Atomics); (LLNL); (Univ. of California at San Diego); (Princeton Plasma Physics Lab.); (Colorado State Univ.); (ParaTools Inc.); (Lehigh Univ.); (ORNL)

2008-01-01

FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

First results from core-edge parallel composition in the FACETS project

Energy Technology Data Exchange (ETDEWEB)

Cary, J R; Carlsson, J A; Hakim, A H; Hamill, P; Kruger, S; Miah, M; Muzsala, S; Pletzer, A; Shasharina, S; Wade-Stein, D; Wang, N [Tech-X Corporation, Boulder, CO 80303 (United States); Candy, J [General Atomics, San Diego, CA 92186 (United States); Cohen, R H [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Krasheninnikov, S [University of California at San Diego, San Diego, CA 92093 (United States); McCune, D C [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Estep, D J [Colorado State University, Fort Collins, CO 80523 (United States); Larson, J [Argonne National Laboratory, Argonne, IL 60439 (United States); Malony, A D [ParaTools, Inc., Eugene, OR 97405 (United States); Pankin, A [Lehigh University, Bethlehem, PA 18015 (United States); Worley, P H [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)], E-mail: cary@txcorp.com (and others)

2008-07-15

FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

First results from core-edge parallel composition in the FACETS project

Energy Technology Data Exchange (ETDEWEB)

Cary, John R. [Tech-X Corporation; Candy, Jeff [General Atomics; Cohen, Ronald H. [Lawrence Livermore National Laboratory (LLNL); Krasheninnikov, Sergei [University of California, San Diego; McCune, Douglas [Princeton Plasma Physics Laboratory (PPPL); Estep, Donald J [Colorado State University, Fort Collins; Larson, Jay [Argonne National Laboratory (ANL); Malony, Allen [University of Oregon; Pankin, A. [Lehigh University, Bethlehem, PA; Worley, Patrick H [ORNL; Carlsson, Johann [Tech-X Corporation; Hakim, A H [Tech-X Corporation; Hamill, P [Tech-X Corporation; Kruger, Scott [Tech-X Corporation; Miah, Mahmood [Tech-X Corporation; Muzsala, S [Tech-X Corporation; Pletzer, Alexander [Tech-X Corporation; Shasharina, Svetlana [Tech-X Corporation; Wade-Stein, D [Tech-X Corporation; Wang, N [Tech-X Corporation; Balay, Satish [Argonne National Laboratory (ANL); McInnes, Lois [Argonne National Laboratory (ANL); Zhang, Hong [Argonne National Laboratory (ANL); Casper, T. A. [Lawrence Livermore National Laboratory (LLNL); Diachin, Lori [Lawrence Livermore National Laboratory (LLNL); Epperly, Thomas [Lawrence Livermore National Laboratory (LLNL); Rognlien, T. D. [Lawrence Livermore National Laboratory (LLNL); Fahey, Mark R [ORNL; Cobb, John W [ORNL; Morris, A [University of Oregon; Shende, Sameer [University of Oregon; Hammett, Greg [Princeton Plasma Physics Laboratory (PPPL); Indireshkumar, K [Tech-X Corporation; Stotler, D. [Princeton Plasma Physics Laboratory (PPPL); Pigarov, A [University of California, San Diego

2008-01-01

FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

Thermohydraulic model of WWER-1000 core

International Nuclear Information System (INIS)

Maroti, L.; Szabados, L.

1987-11-01

Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs

Reactor physics needs in developing countries

International Nuclear Information System (INIS)

Solanilla, R.

1980-01-01

The aim of this paper the identification of needs on Reactor Physics in developing countries embarked in the installation and later on in the operation of Commercial Nuclear Power Plants. In this context the main task of Reactor Physics should be focused in the application of Physical models with inclusion of thermohydraulic process to solve the various realistic problems which appear to ensure a safe, economical and reliable core design and reactor operation. The first part of the paper deals with the scope of Reactor Physics and its interrelation with other disciplines as seen from the view point of developing countries possibilities. Needs requiring a quick response, i.e., those demands coming during the development of a specific Nuclear Power Plant Project, are summarized in the second part of the lecture. Plant startup has been chosen as reference to separate two categories of requirements: Requirements prior to startup phase include reactor core verification, licensing aspects review and study of fuel utilization alternatives; whereas the period during and after startup mainly embraces codes checkup and normalization, core follow-up and long term prediction

Core cooling systems

International Nuclear Information System (INIS)

Hoeppner, G.

1980-01-01

The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

Iron snow in the Martian core?

Science.gov (United States)

Davies, Christopher J.; Pommier, Anne

2018-01-01

The decline of Mars' global magnetic field some 3.8-4.1 billion years ago is thought to reflect the demise of the dynamo that operated in its liquid core. The dynamo was probably powered by planetary cooling and so its termination is intimately tied to the thermochemical evolution and present-day physical state of the Martian core. Bottom-up growth of a solid inner core, the crystallization regime for Earth's core, has been found to produce a long-lived dynamo leading to the suggestion that the Martian core remains entirely liquid to this day. Motivated by the experimentally-determined increase in the Fe-S liquidus temperature with decreasing pressure at Martian core conditions, we investigate whether Mars' core could crystallize from the top down. We focus on the "iron snow" regime, where newly-formed solid consists of pure Fe and is therefore heavier than the liquid. We derive global energy and entropy equations that describe the long-timescale thermal and magnetic history of the core from a general theory for two-phase, two-component liquid mixtures, assuming that the snow zone is in phase equilibrium and that all solid falls out of the layer and remelts at each timestep. Formation of snow zones occurs for a wide range of interior and thermal properties and depends critically on the initial sulfur concentration, ξ0. Release of gravitational energy and latent heat during growth of the snow zone do not generate sufficient entropy to restart the dynamo unless the snow zone occupies at least 400 km of the core. Snow zones can be 1.5-2 Gyrs old, though thermal stratification of the uppermost core, not included in our model, likely delays onset. Models that match the available magnetic and geodetic constraints have ξ0 ≈ 10% and snow zones that occupy approximately the top 100 km of the present-day Martian core.

PWR core and spent fuel pool analysis using scale and nestle

International Nuclear Information System (INIS)

Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

2012-01-01

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

PWR core and spent fuel pool analysis using scale and nestle

Energy Technology Data Exchange (ETDEWEB)

Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

2012-07-01

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

Remanent magnetization stratigraphy of lunar cores

Science.gov (United States)

Banerjee, S. K.; Gingrich, D.; Marvin, J. A.

1977-01-01

Depth dependent fluctuations have been observed in the natural remanent magnetizations (NRM) of drive cores and drill strings from Apollo 16 and 17 missions. Partial demagnetization of unstable secondary magnetizations and identification of characteristic error signals from a core which is known to have been recently disturbed allow us to identify and isolate the stable NRM stratigraphy in double drive core 60010/60009 and drill strings 60002-60004. The observed magnetization fluctuations persist after normalization to take into account depth dependent variations in the carriers of stable NRM. We tentatively ascribe the stable NRM stratigraphy to instantaneous records of past magnetic fields at the lunar surface and suggest that the stable NRM stratigraphy technique could develop as a new relative time-stratigraphic tool, to be used with other physical measurements such as relative intensity of ferromagnetic resonance and charged particle track density to study the evolution of the lunar regolith.

Influence green sand system by core sand additions

Directory of Open Access Journals (Sweden)

N. Špirutová

2012-01-01

Full Text Available Today, about two thirds of iron alloys casting (especially for graphitizing alloys of iron are produced into green sand systems with usually organically bonded cores. Separation of core sands from the green sand mixture is very difficult, after pouring. The core sand concentration increase due to circulation of green sand mixture in a closed circulation system. Furthermore in some foundries, core sands have been adding to green sand systems as a replacement for new sands. The goal of this contribution is: “How the green sand systems are influenced by core sands?”This effect is considered by determination of selected technological properties and degree of green sand system re-bonding. From the studies, which have been published yet, there is not consistent opinion on influence of core sand dilution on green sand system properties. In order to simulation of the effect of core sands on the technological properties of green sands, there were applied the most common used technologies of cores production, which are based on bonding with phenolic resin. Core sand concentration added to green sand system, was up to 50 %. Influence of core sand dilution on basic properties of green sand systems was determined by evaluation of basic industrial properties: moisture, green compression strength and splitting strength, wet tensile strength, mixture stability against staling and physical-chemistry properties (pH, conductivity, and loss of ignition. Ratio of active betonite by Methylene blue test was also determined.

Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

International Nuclear Information System (INIS)

Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

2015-01-01

The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

Assessment of CANDU physics codes using experimental data - II: CANDU core physics measurements

International Nuclear Information System (INIS)

Roh, Gyu Hong; Jeong, Chang Joon; Choi, Hang Bok

2001-11-01

Benchmark calculations of the advanced CANDU reactor analysis tools (WIMS-AECL, SHETAN and RFSP) and the Monte Carlo code MCNP-4B have been performed using Wolsong Units 2 and 3 Phase-B measurement data. In this study, the benchmark calculations have been done for the criticality, boron worth, reactivity device worth, reactivity coefficient, and flux scan. For the validation of the WIMS-AECL/SHETANRFSP code system, the lattice parameters of the fuel channel were generated by the WIMS-AECL code, and incremental cross sections of reactivity devices and structural material were generated by the SHETAN code. The results have shown that the criticality is under-predicted by -4 mk. The reactivity device worths are generally consistent with the measured data except for the strong absorbers such as shutoff rod and mechanical control absorber. The heat transport system temperature coefficient and flux distributions are in good agreement with the measured data. However, the moderator temperature coefficient has shown a relatively large error, which could be caused by the incremental cross-section generation methodology for the reactivity device. For the MCNP-4B benchmark calculation, cross section libraries were newly generated from ENDF/B-VI release 3 through the NJOY97.114 data processing system and a three-dimensional full core model was developed. The simulation results have shown that the criticality is estimated within 4 mk and the estimated reactivity worth of the control devices are generally consistent with the measurement data, which implies that the MCNP code is valid for CANDU core analysis. In the future, therefore, the MCNP code could be used as a reference tool to benchmark design and analysis codes for the advanced fuels for which experimental data are not available

Reactor core operation management system

International Nuclear Information System (INIS)

Sato, Tomomi.

1992-01-01

Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

Reactor core operation management system

Energy Technology Data Exchange (ETDEWEB)

Sato, Tomomi.

1992-05-28

Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

Reactor physics calculations in the Nordic countries

International Nuclear Information System (INIS)

Hoeglund, R.

1995-01-01

The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

Korea's choice of a new generation of nuclear plants

International Nuclear Information System (INIS)

Redding, J.R.

1994-01-01

The ABWR and SBWR design, both under development at GE, provide the best platform for developing the next generation advanced plants. The ABWR, which is rapidly setting the standard for new nuclear reactor plants, is clearly the best choice to meet the present energy needs of Korea. And through a GE/Korea partnership to develop the plant of the next century, Korea will establish itself as a leader in innovative reactor technology

The American Society for Radiation Oncology's 2010 Core Physics Curriculum for Radiation Oncology Residents

Energy Technology Data Exchange (ETDEWEB)

Xiao Ying, E-mail: ying.xiao@jefferson.edu [Thomas Jefferson University Hospital, Philadelphia, PA (United States); De Amorim Bernstein, Karen [Montefiore Medical Center, Bronx, NY (United States); Chetty, Indrin J. [Henry Ford Health System, Detroit, MI (United States); Eifel, Patricia [M. D. Anderson Cancer Center, Houston, TX (United States); Hughes, Lesley [Cooper University Hospital, Camden, NJ (United States); Klein, Eric E. [Washington University, Saint Louis, MO (United States); McDermott, Patrick [William Beaumont Hospital, Royal Oak, MI (United States); Prisciandaro, Joann [University of Michigan, Ann Arbor, MI (United States); Paliwal, Bhudatt [University of Wisconsin, Madison, WI (United States); Price, Robert A. [Fox Chase Cancer Center, Philadelphia, PA (United States); Werner-Wasik, Maria [Thomas Jefferson University Hospital, Philadelphia, PA (United States); Palta, Jatinder R. [University of Florida, Gainesville, FL (United States)

2011-11-15

Purpose: In 2004, the American Society for Radiation Oncology (ASTRO) published its first physics education curriculum for residents, which was updated in 2007. A committee composed of physicists and physicians from various residency program teaching institutions was reconvened again to update the curriculum in 2009. Methods and Materials: Members of this committee have associations with ASTRO, the American Association of Physicists in Medicine, the Association of Residents in Radiation Oncology, the American Board of Radiology (ABR), and the American College of Radiology. Members reviewed and updated assigned subjects from the last curriculum. The updated curriculum was carefully reviewed by a representative from the ABR and other physics and clinical experts. Results: The new curriculum resulted in a recommended 56-h course, excluding initial orientation. Learning objectives are provided for each subject area, and a detailed outline of material to be covered is given for each lecture hour. Some recent changes in the curriculum include the addition of Radiation Incidents and Bioterrorism Response Training as a subject and updates that reflect new treatment techniques and modalities in a number of core subjects. The new curriculum was approved by the ASTRO board in April 2010. We anticipate that physicists will use this curriculum for structuring their teaching programs, and subsequently the ABR will adopt this educational program for its written examination. Currently, the American College of Radiology uses the ASTRO curriculum for their training examination topics. In addition to the curriculum, the committee updated suggested references and the glossary. Conclusions: The ASTRO physics education curriculum for radiation oncology residents has been updated. To ensure continued commitment to a current and relevant curriculum, the subject matter will be updated again in 2 years.

Physical properties of core-concrete systems: Al{sub 2}O{sub 3}-ZrO{sub 2} molten materials measured by aerodynamic levitation

Energy Technology Data Exchange (ETDEWEB)

Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kargl, F. [Institute of Materials Physics in Space, German Aerospace Center (Germany); Nakamori, F.; Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

2017-04-15

During a molten core–concrete interaction, molten oxides consisting of molten core materials (UO{sub 2} and ZrO{sub 2}) and concrete (Al{sub 2}O{sub 3}, SiO{sub 2}, CaO) are formed. Reliable data on the physical properties of the molten oxides will allow us to accurately predict the progression of a nuclear reactor core meltdown accident. In this study, the viscosities and densities of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) were measured using an aerodynamic levitation technique. The densities of two small samples were estimated from their masses and their volumes (calculated from recorded images of the molten samples). The droplets were forced to oscillate using speakers, and their viscosities were evaluated from the damping behaviors of their oscillations. The results showed that the viscosity of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} compared to that of pure molten Al{sub 2}O{sub 3} is 25% lower for x = 0.172, while it is unexpectedly 20% higher for x = 0.356. - Highlights: •The physical properties of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) have been evaluated. •The measurement was conducted using an aerodynamic levitation technique. •The density and viscosity were measured.

Reactors physics. Bases of nuclear physics

International Nuclear Information System (INIS)

Diop, Ch.M.

2006-01-01

The aim of nuclear reactor physics is to quantify the relevant macroscopic data for the characterization of the neutronic state of a reactor core and to evaluate the effects of radiations (neutrons and gamma radiations) on organic matter and on inorganic materials. This first article presents the bases of nuclear physics in the context of nuclear reactors: 1 - reactor physics and nuclear physics; 2 - atomic nucleus - basic definitions: nucleus constituents, dimensions and mass of the atomic nucleus, mass defect, binding energy and stability of the nucleus, strong interaction, nuclear momentums of nucleons and nucleus; 3 - nucleus stability and radioactivity: equation of evolution with time - radioactive decay law; alpha decay, stability limit of spontaneous fission, beta decay, electronic capture, gamma emission, internal conversion, radioactivity, two-body problem and notion of radioactive equilibrium. (J.S.)

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

2014-07-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Liu, Z.

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Solid and gaseous inclusions in the EDML deep ice core: origins and implications for the physical properties of polar ice

Science.gov (United States)

Faria, S. H.; Kipfstuhl, S.; Garbe, C. S.; Bendel, V.; Weikusat, C.; Weikusat, I.

2010-12-01

The great value of polar deep ice cores stems mainly from two essential features of polar ice: its crystalline structure and its impurities. They determine the physical properties of the ice matrix and provide proxies for the investigation of past climates. Experience shows that these two essential features of polar ice manifest themselves in a multiscale diversity of dynamic structures, including dislocations, grain boundaries, solid particles, air bubbles, clathrate hydrates and cloudy bands, among others. The fact that these structures are dynamic implies that they evolve with time through intricate interactions between the crystalline structure, impurities, and the ice flow. Records of these interactions have been carefully investigated in samples of the EPICA deep ice core drilled in Dronning Maud Land, Antarctica (75°S, 0°E, 2882 m elevation, 2774.15 m core length). Here we show how the distributions of sizes and shapes of air bubbles correlate with impurities and the crystalline structure, how the interaction between moving grain boundaries and micro-inclusions changes with ice depth and temperature, as well as the possible causes for the abrupt change in ice rheology observed in the MIS6-MIS5e transition. We also discuss how these observations may affect the flow of the ice sheet and the interpretation of paleoclimate records. Micrograph of an EDML sample from 555m depth. One can identify air bubbles (dark, round objects), microinclusions (tiny defocused spots), and a grain boundary pinned by a bubble. The width of the image is 700 micrometers.

Core power capability verification for PWR NPP

International Nuclear Information System (INIS)

Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

2002-01-01

The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

Bent-core mesogens with an aromatic unit at the terminal position

Czech Academy of Sciences Publication Activity Database

Bajzikova, K.; Svoboda, J.; Novotná, Vladimíra; Pociecha, D.; Gorecka, E.

2017-01-01

Roč. 41, č. 11 (2017), s. 4672-4679 ISSN 1144-0546 R&D Projects: GA ČR GA16-12150S; GA ČR GA15-02843S Institutional support: RVO:68378271 Keywords : bent- core mesogens * liquid crystals Subject RIV: BM - Solid Matter Physics ; Magnetism OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.) Impact factor: 3.269, year: 2016

Real-time simulation of ex-core nuclear instrumentation system

International Nuclear Information System (INIS)

Zhao Qiang; Zhang Zhijian; Cao Xinrong

2005-01-01

Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant (NPP) full-scope training simulator. The simulation method, which is based upon the theory of measurement, is introduced in the paper. The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established. The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration. The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator. (authors)

MO-E-18C-04: Advanced Computer Simulation and Visualization Tools for Enhanced Understanding of Core Medical Physics Concepts

International Nuclear Information System (INIS)

Naqvi, S

2014-01-01

Purpose: Most medical physics programs emphasize proficiency in routine clinical calculations and QA. The formulaic aspect of these calculations and prescriptive nature of measurement protocols obviate the need to frequently apply basic physical principles, which, therefore, gradually decay away from memory. E.g. few students appreciate the role of electron transport in photon dose, making it difficult to understand key concepts such as dose buildup, electronic disequilibrium effects and Bragg-Gray theory. These conceptual deficiencies manifest when the physicist encounters a new system, requiring knowledge beyond routine activities. Methods: Two interactive computer simulation tools are developed to facilitate deeper learning of physical principles. One is a Monte Carlo code written with a strong educational aspect. The code can “label” regions and interactions to highlight specific aspects of the physics, e.g., certain regions can be designated as “starters” or “crossers,” and any interaction type can be turned on and off. Full 3D tracks with specific portions highlighted further enhance the visualization of radiation transport problems. The second code calculates and displays trajectories of a collection electrons under arbitrary space/time dependent Lorentz force using relativistic kinematics. Results: Using the Monte Carlo code, the student can interactively study photon and electron transport through visualization of dose components, particle tracks, and interaction types. The code can, for instance, be used to study kerma-dose relationship, explore electronic disequilibrium near interfaces, or visualize kernels by using interaction forcing. The electromagnetic simulator enables the student to explore accelerating mechanisms and particle optics in devices such as cyclotrons and linacs. Conclusion: The proposed tools are designed to enhance understanding of abstract concepts by highlighting various aspects of the physics. The simulations serve as

MO-E-18C-04: Advanced Computer Simulation and Visualization Tools for Enhanced Understanding of Core Medical Physics Concepts

Energy Technology Data Exchange (ETDEWEB)

Naqvi, S [Saint Agnes Cancer Institute, Department of Radiation Oncology, Baltimore, MD (United States)

2014-06-15

Purpose: Most medical physics programs emphasize proficiency in routine clinical calculations and QA. The formulaic aspect of these calculations and prescriptive nature of measurement protocols obviate the need to frequently apply basic physical principles, which, therefore, gradually decay away from memory. E.g. few students appreciate the role of electron transport in photon dose, making it difficult to understand key concepts such as dose buildup, electronic disequilibrium effects and Bragg-Gray theory. These conceptual deficiencies manifest when the physicist encounters a new system, requiring knowledge beyond routine activities. Methods: Two interactive computer simulation tools are developed to facilitate deeper learning of physical principles. One is a Monte Carlo code written with a strong educational aspect. The code can “label” regions and interactions to highlight specific aspects of the physics, e.g., certain regions can be designated as “starters” or “crossers,” and any interaction type can be turned on and off. Full 3D tracks with specific portions highlighted further enhance the visualization of radiation transport problems. The second code calculates and displays trajectories of a collection electrons under arbitrary space/time dependent Lorentz force using relativistic kinematics. Results: Using the Monte Carlo code, the student can interactively study photon and electron transport through visualization of dose components, particle tracks, and interaction types. The code can, for instance, be used to study kerma-dose relationship, explore electronic disequilibrium near interfaces, or visualize kernels by using interaction forcing. The electromagnetic simulator enables the student to explore accelerating mechanisms and particle optics in devices such as cyclotrons and linacs. Conclusion: The proposed tools are designed to enhance understanding of abstract concepts by highlighting various aspects of the physics. The simulations serve as

Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

Energy Technology Data Exchange (ETDEWEB)

David W. Nigg; Sean R. Morrell

2012-09-01

Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2009-01-01

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2008-01-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

Thermal-hydraulic characteristics of double flat core HCLWR

International Nuclear Information System (INIS)

Sugimoto, Jun; Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio

1989-02-01

A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the same time, high conversion ratio and high burnup can be obtainable. The proposed double flat core HCLWR, based on these physical advantages and the consideration of safety assurance, aims at efficient use of the pressure vessel space to produce comparable thermal output as current 3-loop PWRs. The present work revealed the following items concerning the thermalhydraulic feasibility of the double flat core HCLWR: (1) Main thermal-hydraulic parameters of the plant can be almost the same as current PWRs, showing the use of PWR standard components without major modifications except in core region. (2) Heat removal from the fuel rod in a steady operational condition has enough margin to the critical heat flux (CHF) limit, which is evaluated with the existing CHF correlations. (3) The calculation by REFLA code shows that the maximum cladding temperature in LOCA-reflood is estimated to be far lower than the licensing criteria. It is therefore considered that the proposed double flat core HCLWR is feasible from the point of thermal-hydraulics. Since the available data base has certain applicational limit to the very short core as the present double flat core HCLWR, further detailed assessment is required. (author)

On the core-mass-shell-luminosity relation for shell-burning stars

International Nuclear Information System (INIS)

Jeffery, C.S.; Saint Andrews Univ.

1988-01-01

Core-mass-shell-luminosity relations for several types of shell-burning star have been calculated using simultaneous differential equations derived from simple homology approximations. The principal objective of obtaining a mass-luminosity relation for helium giants was achieved. This relation gives substantially higher luminosities than the equivalent relation for H-shell stars with core masses greater than 1 solar mass. The algorithm for calculating mass-luminosity relations in this fashion was investigated in detail. Most of the assumptions regarding the physics in the shell do not play a critical role in determining the core-mass-shell-luminosity relation. The behaviour of the core-mass-core-radius relation for a growing degenerate core as a single unique function of mass and growth rate needs to be defined before a single core-mass-shell-luminosity relation for all H-shell stars can be obtained directly from the homology approximations. (author)

Model extension and improvement for simulator-based software safety analysis

Energy Technology Data Exchange (ETDEWEB)

Huang, H.-W. [Department of Engineering and System Science, National Tsing Hua University (NTHU), 101 Section 2 Kuang Fu Road, Hsinchu, Taiwan (China) and Institute of Nuclear Energy Research (INER), No. 1000 Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)]. E-mail: hwhwang@iner.gov.tw; Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University (NTHU), 101 Section 2 Kuang Fu Road, Hsinchu, Taiwan (China); Yih Swu [Department of Computer Science and Information Engineering, Ching Yun University, 229 Chien-Hsin Road, Jung-Li, Taoyuan County 320, Taiwan (China); Chen, M.-H. [Institute of Nuclear Energy Research (INER), No. 1000Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Lin, J.-M. [Taiwan Power Company (TPC), 242 Roosevelt Road, Section 3, Taipei 100, Taiwan (China)

2007-05-15

One of the major concerns when employing digital I and C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I and C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes 'simulator/plant model testing' technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I and C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran-ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then further be applied to hazard analysis for operator/digital I and C interface interaction failure study, and the hardware-in-the-loop fault injection study.

ESFR core optimization and uncertainty studies

International Nuclear Information System (INIS)

Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

2015-01-01

In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR

Computational models of stellar collapse and core-collapse supernovae

International Nuclear Information System (INIS)

Ott, Christian D; O'Connor, Evan; Schnetter, Erik; Loeffler, Frank; Burrows, Adam; Livne, Eli

2009-01-01

Core-collapse supernovae are among Nature's most energetic events. They mark the end of massive star evolution and pollute the interstellar medium with the life-enabling ashes of thermonuclear burning. Despite their importance for the evolution of galaxies and life in the universe, the details of the core-collapse supernova explosion mechanism remain in the dark and pose a daunting computational challenge. We outline the multi-dimensional, multi-scale, and multi-physics nature of the core-collapse supernova problem and discuss computational strategies and requirements for its solution. Specifically, we highlight the axisymmetric (2D) radiation-MHD code VULCAN/2D and present results obtained from the first full-2D angle-dependent neutrino radiation-hydrodynamics simulations of the post-core-bounce supernova evolution. We then go on to discuss the new code Zelmani which is based on the open-source HPC Cactus framework and provides a scalable AMR approach for 3D fully general-relativistic modeling of stellar collapse, core-collapse supernovae and black hole formation on current and future massively-parallel HPC systems. We show Zelmani's scaling properties to more than 16,000 compute cores and discuss first 3D general-relativistic core-collapse results.

Tank 101-SY Window E core sample: Interpretation of results

International Nuclear Information System (INIS)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ''Window E.'' This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150 degrees C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts

Tank 101-SY Window E core sample: Interpretation of results

Energy Technology Data Exchange (ETDEWEB)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ``Window E.`` This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150{degrees}C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts.

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

Science.gov (United States)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

In-core fuel management practice in HANARO

International Nuclear Information System (INIS)

Kim Hark Rho; Lee Choong Sung; Lee Jo Bok

1997-01-01

KAERI (KOREA Atomic Energy Research Institute) completed the system performance tests for the HANARO (Hi-flux Advanced Neutron Application Research Reactor) on December 1994. Its initial criticality was achieved on February 8, 1995. A variety of the reactor physics experiments were performed in parallel with configuring the first cycle core and now HANARO is in the third cycle operation. The in-core fuel management in HANARO is performed on the following strategy: 1) the cycle length of the equilibrium core is at least 4 week FPDs, 2) the maximum linear heat generation rate should be within the design limit, 3) the reactor should have shutdown margin of 1% Δk/k at minimum, 4) the available thermal flux should satisfy the users' requirements. This paper presents the fuel management practice in HANARO. Section II briefly describes the design feature of the HANARO and the method of analysis follows in section III and section IV describes In-core fuel management practice and the conclusion is remarked in the final section. (author)

Nuclear detectors for in-core power-reactors

International Nuclear Information System (INIS)

Duchene, Jean; Verdant, Robert.

1979-12-01

Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

Cusp-core problem and strong gravitational lensing

International Nuclear Information System (INIS)

Li Nan; Chen Daming

2009-01-01

Cosmological numerical simulations of galaxy formation have led to the cuspy density profile of a pure cold dark matter halo toward the center, which is in sharp contradiction with the observations of the rotation curves of cold dark matter-dominated dwarf and low surface brightness disk galaxies, with the latter tending to favor mass profiles with a flat central core. Many efforts have been devoted to resolving this cusp-core problem in recent years, among them, baryon-cold dark matter interactions are considered to be the main physical mechanisms erasing the cold dark matter (CDM) cusp into a flat core in the centers of all CDM halos. Clearly, baryon-cold dark matter interactions are not customized only for CDM-dominated disk galaxies, but for all types, including giant ellipticals. We first fit the most recent high resolution observations of rotation curves with the Burkert profile, then use the constrained core size-halo mass relation to calculate the lensing frequency, and compare the predicted results with strong lensing observations. Unfortunately, it turns out that the core size constrained from rotation curves of disk galaxies cannot be extrapolated to giant ellipticals. We conclude that, in the standard cosmological paradigm, baryon-cold dark matter interactions are not universal mechanisms for galaxy formation, and therefore, they cannot be true solutions to the cusp-core problem.

Analysis of a multigroup stylized CANDU half-core benchmark

International Nuclear Information System (INIS)

Pounders, Justin M.; Rahnema, Farzad; Serghiuta, Dumitru

2011-01-01

Highlights: → This paper provides a benchmark that is a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. → An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core CANDU benchmark problem. → Reference eigenvalues and selected pin and bundle fission rates are included. → 2-, 4- and 47-group Monte Carlo solutions are compared to analyze homogenization-free transport approximations that result from energy condensation. - Abstract: An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.

Integrating Condensed Matter Physics into a Liberal Arts Physics Curriculum

Science.gov (United States)

Collett, Jeffrey

2008-03-01

The emergence of nanoscale science into the popular consciousness presents an opportunity to attract and retain future condensed matter scientists. We inject nanoscale physics into recruiting activities and into the introductory and the core portions of the curriculum. Laboratory involvement and research opportunity play important roles in maintaining student engagement. We use inexpensive scanning tunneling (STM) and atomic force (AFM) microscopes to introduce students to nanoscale structure early in their college careers. Although the physics of tip-surface interactions is sophisticated, the resulting images can be interpreted intuitively. We use the STM in introductory modern physics to explore quantum tunneling and the properties of electrons at surfaces. An interdisciplinary course in nanoscience and nanotechnology course team-taught with chemists looks at nanoscale phenomena in physics, chemistry, and biology. Core quantum and statistical physics courses look at effects of quantum mechanics and quantum statistics in degenerate systems. An upper level solid-state physics course takes up traditional condensed matter topics from a structural perspective by beginning with a study of both elastic and inelastic scattering of x-rays from crystalline solids and liquid crystals. Students encounter reciprocal space concepts through the analysis of laboratory scattering data and by the development of the scattering theory. The course then examines the importance of scattering processes in band structure and in electrical and thermal conduction. A segment of the course is devoted to surface physics and nanostructures where we explore the effects of restricting particles to two-dimensional surfaces, one-dimensional wires, and zero-dimensional quantum dots.

Test model of WWER core

International Nuclear Information System (INIS)

Tikhomirov, A. V.; Gorokhov, A. K.

2007-01-01

The objective of this paper is creation of precision test model for WWER RP neutron-physics calculations. The model is considered as a tool for verification of deterministic computer codes that enables to reduce conservatism of design calculations and enhance WWER RP competitiveness. Precision calculations were performed using code MCNP5/1/ (Monte Carlo method). Engineering computer package Sapfir 9 5andRC V VER/2/ is used in comparative analysis of the results, it was certified for design calculations of WWER RU neutron-physics characteristic. The object of simulation is the first fuel loading of Volgodon NPP RP. Peculiarities of transition in calculation using MCNP5 from 2D geometry to 3D geometry are shown on the full-scale model. All core components as well as radial and face reflectors, automatic regulation in control and protection system control rod are represented in detail description according to the design. The first stage of application of the model is assessment of accuracy of calculation of the core power. At the second stage control and protection system control rod worth was assessed. Full scale RP representation in calculation using code MCNP5 is time consuming that calls for parallelization of computational problem on multiprocessing computer (Authors)

VHTR core modeling: coupling between neutronic and thermal-hydraulics

International Nuclear Information System (INIS)

Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

2005-01-01

Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

2008-07-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

2009-04-15

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

Multi-core System Architecture for Safety-critical Control Applications

DEFF Research Database (Denmark)

Li, Gang

and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...

Updating the Psoriatic Arthritis (PsA) Core Domain Set

DEFF Research Database (Denmark)

Orbai, Ana-Maria; de Wit, Maarten; Mease, Philip J

2017-01-01

OBJECTIVE: To include the patient perspective in accordance with the Outcome Measures in Rheumatology (OMERACT) Filter 2.0 in the updated Psoriatic Arthritis (PsA) Core Domain Set for randomized controlled trials (RCT) and longitudinal observational studies (LOS). METHODS: At OMERACT 2016, research...... conducted to update the PsA Core Domain Set was presented and discussed in breakout groups. The updated PsA Core Domain Set was voted on and endorsed by OMERACT participants. RESULTS: We conducted a systematic literature review of domains measured in PsA RCT and LOS, and identified 24 domains. We conducted...... and breakout groups at OMERACT 2016 in which findings were presented and discussed. The updated PsA Core Domain Set endorsed with 90% agreement by OMERACT 2016 participants included musculoskeletal disease activity, skin disease activity, fatigue, pain, patient's global assessment, physical function, health...

Development of next BWR plant

International Nuclear Information System (INIS)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

1995-01-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

Development of next BWR plant

Energy Technology Data Exchange (ETDEWEB)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1995-04-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

Reload core safety verification

International Nuclear Information System (INIS)

Svetlik, M.; Minarcin, M.

2003-01-01

This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

Core-shell magnetic nanowires fabrication and characterization

Energy Technology Data Exchange (ETDEWEB)

Kalska-Szostko, B., E-mail: kalska@uwb.edu.pl [Institute of Chemistry, University of Bialystok, Ciolkowskiego 1K, 15-245 Bialystok (Poland); Faculty of Physics, University of Bialystok, Ciolkowskiego 1L, 15-245 Bialystok, Poland (Poland); Klekotka, U.; Satuła, D. [Institute of Chemistry, University of Bialystok, Ciolkowskiego 1K, 15-245 Bialystok (Poland); Faculty of Physics, University of Bialystok, Ciolkowskiego 1L, 15-245 Bialystok, Poland (Poland)

2017-02-28

Highlights: • New approach for nanowires modification are presented. • Physical and chemical characterization of the nanowires are shown. • Properties modulations as an effect of the surface layer composition are discussed. - Abstract: In this paper, a new way of the preparation of core-shell magnetic nanowires has been proposed. For the modification Fe nanowires were prepared by electrodeposition in anodic aluminium oxide matrixes, in first step. In second, by wetting chemical deposition, shell layers of Ag, Au or Cu were obtained. Resultant core-shell nanowires structure was characterized by X-ray diffraction, infrared spectroscopy, transmission electron microscopy, and energy dispersive x-ray. Whereas magnetic properties by Mössbauer spectroscopy.

Core Hunter 3: flexible core subset selection.

Science.gov (United States)

De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

2018-05-31

Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

Physics to a degree

CERN Document Server

Thomas, EG

2014-01-01

Physics to a Degree provides an extensive collection of problems suitable for self-study or tutorial and group work at the level of an undergraduate physics course. This novel set of exercises draws together the core elements of an undergraduate physics degree and provides students with the problem solving skills needed for general physics' examinations and for real-life situations encountered by the professional physicist. Topics include force, momentum, gravitation, Bernoulli's Theorem, magnetic fields, blackbody radiation, relativistic travel, mechanics near the speed of light, radioactive

Nigeria Journal of Pure and Applied Physics

African Journals Online (AJOL)

Nigeria Journal of Pure and Applied Physics publishes papers of the highest quality and significance in specific areas of physics, pure and applied, as listed below. The journal content reflects core physics disciplines, but is also open to a broad range of topics whose central theme falls within the bounds of physics.

Properties of polyurethane foam/coconut coir fiber as a core material and as a sandwich composites component

Science.gov (United States)

Azmi, M. A.; Abdullah, H. Z.; Idris, M. I.

2013-12-01

This research focuses on the fabrication and characterization of sandwich composite panels using glass fiber composite skin and polyurethane foam reinforced coconut coir fiber core. The main objectives are to characterize the physical and mechanical properties and to elucidate the effect of coconut coir fibers in polyurethane foam cores and sandwich composite panels. Coconut coir fibers were used as reinforcement in polyurethane foams in which later were applied as the core in sandwich composites ranged from 5 wt% to 20 wt%. The physical and mechanical properties found to be significant at 5 wt% coconut coir fiber in polyurethane foam cores as well as in sandwich composites. It was found that composites properties serve better in sandwich composites construction.

Properties of polyurethane foam/coconut coir fiber as a core material and as a sandwich composites component

International Nuclear Information System (INIS)

Azmi, M A; Abdullah, H Z; Idris, M I

2013-01-01

This research focuses on the fabrication and characterization of sandwich composite panels using glass fiber composite skin and polyurethane foam reinforced coconut coir fiber core. The main objectives are to characterize the physical and mechanical properties and to elucidate the effect of coconut coir fibers in polyurethane foam cores and sandwich composite panels. Coconut coir fibers were used as reinforcement in polyurethane foams in which later were applied as the core in sandwich composites ranged from 5 wt% to 20 wt%. The physical and mechanical properties found to be significant at 5 wt% coconut coir fiber in polyurethane foam cores as well as in sandwich composites. It was found that composites properties serve better in sandwich composites construction

Development of Core Simulator (CoSi) for APR1400 And Analysis of LPPT Result using APR1400-CoSi

International Nuclear Information System (INIS)

Moon, Sang-Rae; Kim, Yong-Bae; Lee, Eun-Ki

2014-01-01

According to NRC guidelines, Low Power Physics Test (LPPT) is required to be performed in low temperature/low pressure (160 .deg. C/42.2 kg/cm 2 ) as well as NOT/NOP (291.3 .deg. C/158.2 kg/cm 2 ) because Shin-Kori Unit 3 is FOAK nuclear power plant. Low Power Physics Test (LPPT) is essential to verify the nuclear design and robustness of reactor safety. LPPT consists of initial criticality, Point of Adding Heat (POAH), All Rod Out (ARO) Boron Concentration, Isothermal Temperature Coefficient (ITC), Control Rod Worth measurements and so on. KHNP-CRI has developed the Core Simulator for APR1400 (APR1400-CoSi) in order to improve the ability performing the LPPT. Especially, APR1400-CoSi has enhanced capability to calculate the full Core neutronic parameters by revising RAST-K that is three dimensional real time core kinetic program. APR1400-CoSi has been developed for education-training of Low Power Physics Test(LPPT). Particularly, APR1400-CoSi has an enhanced capability to calculate the full core neutronic parameters by revising RAST-K which is a three dimensional real time core kinetics program. Low Power Physics Test (LPPT) was performed using APR1400-CoSi and the results showed very similar values with the predicted ones. In other words, the initial core model of Shin-Kori Unit 3 in APR1400-CoSi system has been verified to be appropriate enough. Also, it was confirmed that the test range of Low Power Physics Test (LPPT) reamains effective even though the largest incremental bank reactivity is inserted in core by analyzing the power change during the rod SWAP test

Nuclear physics, neutron physics and nuclear energy. Proceedings

International Nuclear Information System (INIS)

Andrejtscheff, W.; Elenkov, D.

1994-01-01

The book contains of proceedings of XI International School on Nuclear Physics, Neutron Physics and Nuclear Energy organized traditionally every two years by Bulgarian Academy of Sciences and the Physics Department of Sofia University held near the city of Varna. It provides a good insight to the large range of theoretical and experimental results, prospects, problems, difficulties and challenges which are at the core of nuclear physics today. The efforts and achievements of scientists to search for new phenomena in nuclei at extreme circumstances as superdeformation and band crossing in nuclear structure understanding are widely covered. From this point of view the achievements and future in the field of high-precision γ-spectroscopy are included. Nuclear structure models and methods, models for strong interaction, particle production and properties, resonance theory and its application in reactor physics are comprised also. (V.T.)

Modeling of the core of Atucha II nuclear power plant

International Nuclear Information System (INIS)

Blanco, Anibal

2007-01-01

This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

Mechanical and Physical Properties of Low Density Kenaf Core Particleboards Bonded with Different Resins

OpenAIRE

Mohamad Jani Saad; Izran Kamal

2012-01-01

Single layer kenaf (Hibiscus cannabinus) core particleboards bonded with urea formaldehyde (UF), phenol formaldehyde (PF) and polymeric 4,4-methyl phenylmethane di-isocyanate (PMDI) resins were manufactured. The boards were fabricated with three different densities i.e 350 kg/m3, 450 kg/m3 and 550 kg/m3. Each type of the resin used was sprayed at three different resin loadings on the kenaf core particles. The boards produced was evaluated for its modulus of rupture (MOR), modulus of elasticit...

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

1982-01-01

Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

Computational models of stellar collapse and core-collapse supernovae

Energy Technology Data Exchange (ETDEWEB)

Ott, Christian D; O' Connor, Evan [TAPIR, Mailcode 350-17, California Institute of Technology, Pasadena, CA (United States); Schnetter, Erik; Loeffler, Frank [Center for Computation and Technology, Louisiana State University, Baton Rouge, LA (United States); Burrows, Adam [Department of Astrophysical Sciences, Princeton University, Princeton, NJ (United States); Livne, Eli, E-mail: cott@tapir.caltech.ed [Racah Institute of Physics, Hebrew University, Jerusalem (Israel)

2009-07-01

Core-collapse supernovae are among Nature's most energetic events. They mark the end of massive star evolution and pollute the interstellar medium with the life-enabling ashes of thermonuclear burning. Despite their importance for the evolution of galaxies and life in the universe, the details of the core-collapse supernova explosion mechanism remain in the dark and pose a daunting computational challenge. We outline the multi-dimensional, multi-scale, and multi-physics nature of the core-collapse supernova problem and discuss computational strategies and requirements for its solution. Specifically, we highlight the axisymmetric (2D) radiation-MHD code VULCAN/2D and present results obtained from the first full-2D angle-dependent neutrino radiation-hydrodynamics simulations of the post-core-bounce supernova evolution. We then go on to discuss the new code Zelmani which is based on the open-source HPC Cactus framework and provides a scalable AMR approach for 3D fully general-relativistic modeling of stellar collapse, core-collapse supernovae and black hole formation on current and future massively-parallel HPC systems. We show Zelmani's scaling properties to more than 16,000 compute cores and discuss first 3D general-relativistic core-collapse results.

TURBULENCE DECAY AND CLOUD CORE RELAXATION IN MOLECULAR CLOUDS

International Nuclear Information System (INIS)

Gao, Yang; Law, Chung K.; Xu, Haitao

2015-01-01

The turbulent motion within molecular clouds is a key factor controlling star formation. Turbulence supports molecular cloud cores from evolving to gravitational collapse and hence sets a lower bound on the size of molecular cloud cores in which star formation can occur. On the other hand, without a continuous external energy source maintaining the turbulence, such as in molecular clouds, the turbulence decays with an energy dissipation time comparable to the dynamic timescale of clouds, which could change the size limits obtained from Jean's criterion by assuming constant turbulence intensities. Here we adopt scaling relations of physical variables in decaying turbulence to analyze its specific effects on the formation of stars. We find that the decay of turbulence provides an additional approach for Jeans' criterion to be achieved, after which gravitational infall governs the motion of the cloud core. This epoch of turbulence decay is defined as cloud core relaxation. The existence of cloud core relaxation provides a more complete understanding of the effect of the competition between turbulence and gravity on the dynamics of molecular cloud cores and star formation

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

2017-05-15

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Two approaches to meeting the economic challenge for advanced BWR designs

International Nuclear Information System (INIS)

Rao, A.S.; Sawyer, C.D.

1997-01-01

This paper presents the design overview and approach to addressing the aforementioned economic challenges for two Advanced BWR designs. The first plant is the ABWR and the second is the ESBWR. The ABWR relies on proven technology and components and an extensive infrastructure that has been built up over the last 20 years. Because it has proven and standards safety systems it has very limited uncertainty regarding licensing. Finally, it relies on the economies of scale and overall design flexibility to improve the overall economics of power generation. The ESBWR on the other hand has taken an innovative approach to reduce systems and components to simplify the overall plant to improve plant economics. The overall plant design is indeed simpler, but improved economics required reliance on some economies of scale also. This design embodied in the ESBWR, also has minimized the overall development cost by utilizing features and components from the ABWR and SBWR technology programs

Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

Energy Technology Data Exchange (ETDEWEB)

Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Tautges, Timothy [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, Robert Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-12-21

This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

Geomagnetic Secular Variation and Its Applications to the Core

DEFF Research Database (Denmark)

Jackson, Andrew; Finlay, Chris

2015-01-01

the mathematical models of the field’s evolution that can be derived from them. We discussthe prominent features of the field’s evolution, both at the Earth’s surface and at the surface of theliquid core. The final section concerns itself with a discussion of the interpretation of the field’sevolution, in terms......We review the observational constraints on the morphology and evolution of the magnetic field of the Earth over the last few centuries; these changes are referred to as the secular variation.Starting with a description of the available sources of original observations of the field, we thendiscuss...... of physical core processes. These divide themselves into processes associatedwith movement of core fluid, which is capable of advecting the field, and processes associated withthe finite resistivity of the core, commonly termed diffusive processes. We lay the foundations for some of the more theoretical...

Research on V and V strategy of reactor physics code of COSINE

International Nuclear Information System (INIS)

Liu Zhanquan; Chen Yixue; Yang Chao; Dang Halei

2013-01-01

Verification and validation (V and V) is very important for the software quality assurance. Reasonable and efficient V and V strategy can achieve twice the result with half the effort. Core and system integrated engine for design and analysis (COSINE) software package contains three reactor physics codes, the lattice code (LATC), the core simulator (CORE) and the kinetics code (KIND), which is called the reactor physics subsystem. The V and V strategy for the physics subsystem was researched based on the foundation of scientific software's V and V method. The module based verification method and the function based validation method were proposed, composing the physical subsystem V and V strategy of COSINE software package. (authors)

LWR-core behaviour project

International Nuclear Information System (INIS)

Paratte, J.M.

1982-07-01

The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

CURRICULUM MATTERS: Physics 2000: a personal view

Science.gov (United States)

Field, R. J.

1997-03-01

The author expresses his personal views of how Physics for A-level should develop towards the year 2000. These cover: the historical treatment of core topics, syllabus structure and the relevance of practical physics.

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

Design and analysis of three-layer-core optical fiber

Science.gov (United States)

Zheng, Siwen; Liu, Yazhuo; Chang, Guangjian

2018-03-01

A three-layer-core single-mode large-mode-area fiber is investigated. The three-layer structure in the core, which is composed of a core-index layer, a cladding-index layer, and a depression-index layer, could achieve a large effective area Aeff while maintaining an ultralow bending loss without deteriorating cutoff behaviors. The single-mode large mode area of 100 to 330 μm2 could be achieved in the fiber. The effective area Aeff can be further enlarged by adjusting the layer parameters. Furthermore, the bending property could be improved in this three-layer-core structure. The bending loss could decrease by 2 to 4 orders of magnitude compared with the conventional step-index fiber with the same Aeff. These characteristics of three-layer-core fiber suggest that it can be used in large-mode-area wide-bandwidth high-capacity transmission or high-power optical fiber laser and amplifier in optical communications, which could be used for the basic physical layer structure of big data storage, reading, calculation, and transmission applications.

Core-to-core uniformity improvement in multi-core fiber Bragg gratings

Science.gov (United States)

Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

2014-07-01

Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

ICE CHEMISTRY IN STARLESS MOLECULAR CORES

Energy Technology Data Exchange (ETDEWEB)

Kalvans, J., E-mail: juris.kalvans@venta.lv [Engineering Research Institute “Ventspils International Radio Astronomy Center” of Ventspils University College, Inzenieru 101, Ventspils, LV-3601 (Latvia)

2015-06-20

Starless molecular cores are natural laboratories for interstellar molecular chemistry research. The chemistry of ices in such objects was investigated with a three-phase (gas, surface, and mantle) model. We considered the center part of five starless cores, with their physical conditions derived from observations. The ice chemistry of oxygen, nitrogen, sulfur, and complex organic molecules (COMs) was analyzed. We found that an ice-depth dimension, measured, e.g., in monolayers, is essential for modeling of chemistry in interstellar ices. Particularly, the H{sub 2}O:CO:CO{sub 2}:N{sub 2}:NH{sub 3} ice abundance ratio regulates the production and destruction of minor species. It is suggested that photodesorption during the core-collapse period is responsible for the high abundance of interstellar H{sub 2}O{sub 2} and O{sub 2}H and other species synthesized on the surface. The calculated abundances of COMs in ice were compared to observed gas-phase values. Smaller activation barriers for CO and H{sub 2}CO hydrogenation may help explain the production of a number of COMs. The observed abundance of methyl formate HCOOCH{sub 3} could be reproduced with a 1 kyr, 20 K temperature spike. Possible desorption mechanisms, relevant for COMs, are gas turbulence (ice exposure to interstellar photons) or a weak shock within the cloud core (grain collisions). To reproduce the observed COM abundances with the present 0D model, 1%–10% of ice mass needs to be sublimated. We estimate that the lifetime for starless cores likely does not exceed 1 Myr. Taurus cores are likely to be younger than their counterparts in most other clouds.

THE EVOLUTION OF CLOUD CORES AND THE FORMATION OF STARS

International Nuclear Information System (INIS)

Broderick, Avery E.; Keto, Eric

2010-01-01

For a number of starless cores, self-absorbed molecular line and column density observations have implied the presence of large-amplitude oscillations. We examine the consequences of these oscillations on the evolution of the cores and the interpretation of their observations. We find that the pulsation energy helps support the cores and that the dissipation of this energy can lead toward instability and star formation. In this picture, the core lifetimes are limited by the pulsation-decay timescales, dominated by non-linear mode-mode coupling, and on the order of ≅ few x 10 5 -10 6 yr. Notably, this is similar to what is required to explain the relatively low rate of conversion of cores into stars. For cores with large-amplitude oscillations, dust continuum observations may appear asymmetric or irregular. As a consequence, some of the cores that would be classified as super-critical may be dynamically stable when oscillations are taken into account. Thus, our investigation motivates a simple hydrodynamic picture, capable of reproducing many of the features of the progenitors of stars without the inclusion of additional physical processes, such as large-scale magnetic fields.

THE EVOLUTION OF CLOUD CORES AND THE FORMATION OF STARS

Energy Technology Data Exchange (ETDEWEB)

Broderick, Avery E [Canadian Institute for Theoretical Astrophysics, 60 St. George Street, Toronto, ON M5S 3H8 (Canada); Keto, Eric [Smithsonian Observatory, 60 Garden Street, Cambridge, MA 02138 (United States)

2010-09-20

For a number of starless cores, self-absorbed molecular line and column density observations have implied the presence of large-amplitude oscillations. We examine the consequences of these oscillations on the evolution of the cores and the interpretation of their observations. We find that the pulsation energy helps support the cores and that the dissipation of this energy can lead toward instability and star formation. In this picture, the core lifetimes are limited by the pulsation-decay timescales, dominated by non-linear mode-mode coupling, and on the order of {approx_equal} few x 10{sup 5}-10{sup 6} yr. Notably, this is similar to what is required to explain the relatively low rate of conversion of cores into stars. For cores with large-amplitude oscillations, dust continuum observations may appear asymmetric or irregular. As a consequence, some of the cores that would be classified as super-critical may be dynamically stable when oscillations are taken into account. Thus, our investigation motivates a simple hydrodynamic picture, capable of reproducing many of the features of the progenitors of stars without the inclusion of additional physical processes, such as large-scale magnetic fields.

High-Resolution Physical Properties Logging of the AND-1B Sediment Core - Opportunity for Detecting High-Frequency Signals of Paleoenvironmental Changes

Science.gov (United States)

Niessen, F.; Magens, D.; Kuhn, G.; Helling, D.

2008-12-01

Within the ANDRILL-MIS Project, a more than 1200 m long sediment core, dating back to about 13 Ma, was drilled beneath McMurdo Ice Shelf near Ross Island (Antarctica) in austral summer 2006/07 with the purpose of contributing to a better understanding of the Late Cenozoic history of the Antarctic Ice Sheet. One way to approach past ice dynamics and changes in the paleoenvironment quantitatively, is the analysis of high- resolution physical properties obtained from whole-core multi-sensor core logger measurements in which lithologic changes are expressed numerically. This is especially applicable for the repeating sequences of diatomites and diamictites in the upper half of the core with a prominent cyclicity between 140-300 mbsf. Rather abrupt high-amplitude variations in wet-bulk density (WBD) and magnetic susceptibility (MS) reflect a highly dynamic depositional system, oscillating between two main end-member types: a grounded ice sheet and open marine conditions. For the whole core, the WBD signal, ranging from 1.4 kg/cu.m in the diatomites to 2.3 kg/cu.m in diamictites from the lower part of the core, represents the influence of three variables: (i) the degree of compaction seen as reduction of porosities with depth of about 30 % from top to bottom, (ii) the clast content with clasts being almost absent in diatomite deposits and (iii) the individual grain density (GD). GD itself strongly reflects the variety of lithologies as well as the influence of cement (mainly pyrite and carbonate) on the matrix grain density. The calculation of residual porosities demonstrates the strong imprint of glacial loading for especially diamictites from the upper 150 m, pointing to a significant thickness of the overriding Pleistocene ice sheet. MS on the other hand mainly documents a marine vs. terrestrial source of sediments where the latter can be divided into younger local material from the McMurdo Volcanic Province and basement clasts from the Transantarctic Mountains

Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

Energy Technology Data Exchange (ETDEWEB)

Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

2016-08-15

Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

International Nuclear Information System (INIS)

Fabbris, Olivier; Dardour, Saied; Blaise, Patrick; Ferrasse, Jean-Henry; Saez, Manuel

2016-01-01

Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

Nuclear reactor core modelling in multifunctional simulators

International Nuclear Information System (INIS)

Puska, E.K.

1999-01-01

The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

Nuclear reactor core modelling in multifunctional simulators

Energy Technology Data Exchange (ETDEWEB)

Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

1999-06-01

The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

The effect of short-term isometric training on core/torso stiffness.

Science.gov (United States)

Lee, Benjamin; McGill, Stuart

2017-09-01

"Core" exercise is a basic part of many physical training regimens with goals ranging from rehabilitation of spine and knee injuries to improving athletic performance. Core stiffness has been proposed to perform several functions including reducing pain by minimising joint micro-movements, and enhancing strength and speed performance. This study probes the links between a training approach and immediate but temporary changes in stiffness. Passive and active stiffness was measured on 24 participants; 12 having little to no experience in core training (inexperienced), and the other 12 being athletes experienced to core training methods; before and after a 15 min bout of isometric core exercises. Passive stiffness was assessed on a "frictionless" bending apparatus and active stiffness assessed via a quick release mechanism. Short-term isometric core training increased passive and active stiffness in most directions for both inexperienced and experienced participants, passive left lateral bend among experienced participants being the exception (P core stiffness, in this case following a single session. This may influence performance and injury resilience for a brief period.

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Sodium, Iodine and Bromine in Polar Ice Cores

DEFF Research Database (Denmark)

Maffezzoli, Niccolo

Abstract: This research focuses on sodium, bromine and iodine in polar ice cores, with the aim of reviewing and advancing their current understanding with additional measurements and records, and investigating the connections of these tracers with sea ice and their feasibility as sea ice indicators...... with a description of the main analytic al techniques used to measure ionic and elemental species in ice cores. Chapter 4 introduces sodium, bromine and iodine with a theoretical perspective and a particular focus on their connections with sea ice. Some of the physical and chemical properties that are believed...... back trajectory analyses of the past 17 years. The results identify the aerosol source area influencing the Renland ice cap, a result necessary for the interpretation of impurity records obtained from the ice core. Chapter 6 reviews the published ice/snow measurements of bromine and iodine at polar...

Features of formation of personal physical education of students

Directory of Open Access Journals (Sweden)

V.A. Sutula

2013-08-01

Full Text Available The peculiarities of formation of personal physical training of students from various universities. The experiment was attended by 748 students. It is shown that the most significant factor in personal physical training of students is to conduct a healthy physically active lifestyle and relationship to the body as a value. It is confirmed that for students of physical culture the main motivational factors for physical culture and sports activities is the desire to improve physical fitness, achieve sporting success. For students of non-core specialties - to optimize the weight and improve the shape. It is proved that the main obstacle to employment physical and sports activities of the students is the lack of free time. It is shown that only 21.2% of students believe their non-core specialties: motor sufficient for normal life and preservation of health.

Final Report for the "Fusion Application for Core-Edge Transport Simulations (FACETS)"

Energy Technology Data Exchange (ETDEWEB)

Cary, John R; Kruger, Scott

2014-10-02

The FACETS project over its lifetime developed the first self-consistent core-edge coupled capabilities, a new transport solver for modeling core transport in tokamak cores, developed a new code for modeling wall physics over long time scales, and significantly improved the capabilities and performance of legacy components, UEDGE, NUBEAM, GLF23, GYRO, and BOUT++. These improved capabilities leveraged the team’s expertise in applied mathematics (solvers and algorithms) and computer science (performance improvements and language interoperability). The project pioneered new methods for tackling the complexity of simulating the concomitant complexity of tokamak experiments.

ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA

International Nuclear Information System (INIS)

1978-01-01

1 - Nature of the physical problem solved: ASCOT-1 is used to analyze the thermo-hydraulic behaviour in a PWR core during a loss-of-coolant accident. 2 - Method of solution: The core is assumed to be axisymmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of fuel in the annular regions into which the core is divided, the heat conduction equations are solved by an explicit method with averaged flow conditions. 3 - Restrictions on the complexity of the problem: Axisymmetric two-dimensional homogeneous flows

Coupling between core and cladding modes in a helical core fiber with large core offset

International Nuclear Information System (INIS)

Napiorkowski, Maciej; Urbanczyk, Waclaw

2016-01-01

We analyzed the effect of resonant coupling between core and cladding modes in a helical core fiber with large core offset using the fully vectorial method based on the transformation optics formalism. Our study revealed that the resonant couplings to lower order cladding modes predicted by perturbative methods and observed experimentally in fibers with small core offsets are in fact prohibited for larger core offsets. This effect is related to the lack of phase matching caused by elongation of the optical path of the fundamental modes in the helical core. Moreover, strong couplings to the cladding modes of the azimuthal modal number much higher than predicted by perturbative methods may be observed for large core offsets, as the core offset introduces higher order angular harmonics in the field distribution of the fundamental modes. Finally, in contrast to previous studies, we demonstrate the existence of spectrally broad polarization sensitive couplings to the cladding modes suggesting that helical core fibers with large core offsets may be used as broadband circular polarizers. (paper)

Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

International Nuclear Information System (INIS)

S, Tukiran; MS, Tagor; P, Surian

2003-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

Self-assessment of professionalism in physical therapy education.

Science.gov (United States)

Anderson, Deborah K; Irwin, Kent E

2013-01-01

With the physical therapy (PT) professions' advancement to the clinical doctorate degree and the promotion of autonomous practice, exemplary professional conduct is an expectation of the PT profession. PT education programs are being challenged to develop methods to teach and assess professional behavior. Forty-three PT students (11 male and 32 female, ages 20-28 years) completed the APTA Professionalism in Physical Therapy: Core Values Self-Assessment (PPTCVSA) after their first 3 week clinical experience and again after their final clinical experience. A mixed design ANOVA compared participants' total scores and individual Core Value scores on the Professionalism in Physical Therapy: Core Values Self-Assessment (PPTCVSA) after 3 and 33 weeks of clinical education. The effects of gender, age, and undergraduate area of study on growth in professionalism scores were also investigated. Total PPTCVSA scores and individual Core Value scores on professionalism (accountability, altruism, compassion/caring, excellence, integrity, professional duty, and social responsibility) were higher after 33 weeks compared to scores after 3 weeks of clinical education. Female student's total professionalism scores were higher than male student's scores on both the first and second self-assessments. In addition, female students scored themselves higher than their male peers on accountability, excellence, integrity, and professional duty. Improved scores on the PPTCVSA indicate that physical therapy education is playing an important role in the development of professional behavior, knowledge, and application in practice.

A simulated test of physical starting and reactor physics on zero power facility of PWR

International Nuclear Information System (INIS)

Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

1995-01-01

The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

Mineral Physics Research on Earth's Core and UTeach Outreach Activities at UT Austin

Science.gov (United States)

Lin, J.; Wheat, A. J.

2011-12-01

Comprehension of the alloying effects of major candidate light elements on the phase diagram and elasticity of iron addresses pressing issues on the composition, thermal structures, and seismic features of the Earth's core. Integrating this mineral physics research with the educational objectives of the CAREER award was facilitated by collaboration with the University of Texas at Austin's premier teaching program, UTeach. The UTeach summer outreach program hosts three one-week summer camps every year exposing K-12th graders to university level academia, emphasizing math and science initiatives and research. Each week of the camp either focuses on math, chemistry, or geology. Many of the students were underrepresented minorities and some required simultaneous translation; this is an effect of the demographics of the region, and caused some language barrier challenges. The students' opportunity to see first-hand what it is like to be on a university campus, as well as being in a research environment, such as the mineral physics lab, helps them to visualize themselves in academia in the future. A collection of displayable materials with information about deep-Earth research were made available to participating students and teachers to disseminate accurate scientific knowledge and enthusiasm. These items included a diamond anvil cell and diagrams of the diamond crystal structure, the layers of the Earth, and the phases of carbon to show that one element can have very different physical properties purely based on differences in structure. The students learned how advanced X-ray and optical laser spectroscopies are used to study properties of planetary materials in the diamond anvil cell. Stress was greatly placed on the basic mathematical relationship between force, area, and pressure, the fundamental principle involved with diamond anvil cell research. Undergraduate researchers from the lab participated in the presentations and hands-on experiments, and answered any

Development of high conversion boiling water reactor

International Nuclear Information System (INIS)

Yamashita, Jun-ichi; Mochida, Takaaki; Uchikawa, Sadao.

1988-01-01

It is expected that the period of LWRs being the main source of electric power supply becomes long, therefore, the development of next generation LWRs placing emphasis on the effective utilization of uranium resources and the improvement of economical efficiency is necessary. In this paper, as the next generation BWRs subsequent to ABWRs, the concept of the core of high conversion type BWRs is reported, in which emphasis is placed on the saving of natural uranium resources by raising the rate of conversion to plutonium. This core is that of realizing the high rate of conversion by utilizing the void in the core, which is the feature of BWRs, and the case of making the change of the core structure relatively small by using cross type control rods and the case of changing the core structure for further heightening the rate of conversion and making control rods into cluster type are described. In order to meet the demand like this, Hitachi Ltd. has engaged in the development of the concept of the core of next generation LWRs. In the high conversion type BWRs, there is not large change in the reactor system and turbine system from the current BWRs. The features and the concept of the core of high conversion type BWRs are described. (Kako, I.)

Earth's inner core: Innermost inner core or hemispherical variations?

NARCIS (Netherlands)

Lythgoe, K. H.; Deuss, A.|info:eu-repo/dai/nl/412396610; Rudge, J. F.; Neufeld, J. A.

2014-01-01

The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

Dependence of Core and Extended Flux on Core Dominance ...

Indian Academy of Sciences (India)

Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

The simultaneous measurements of core and outer core density fluctuations in L-H transition using CO2 laser collective scattering diagnostic in the EAST superconducting tokamak

International Nuclear Information System (INIS)

Cao, G.M.; Li, Y.D.; Zhang, X.D.; Sun, P.J.; Hu, L.Q.; Li, J.G.; Wu, G.J.

2013-01-01

The H-mode is the projected basic operation scenario for the ITER tokamak. The turbulence de-correlation by the synergistic effect of zonal flow and equilibrium ExB flow shear is believed to be the reason for L-H transition, however, the detailed physical mechanism has not been identified so far. Tangential multi-channel CO 2 laser collective scattering diagnostic system (mainly k r measurement) was first installed to investigate electron density fluctuations on EAST tokamak. The measurements in a spontaneous dithering L-H transition show that in core plasma (0 < r/a < 0.5) the low-frequency fluctuations strengthen greatly before L-H transition; meanwhile in outer core plasma (0.2 < r/a < 1) the low-frequency fluctuations strengthen slightly. Bispectral analysis reveals that the coupling strength between low- and high-frequency fluctuations in both core and outer core plasma strengthens greatly before the transition, but the latter is greater than the former. The results indicate that the low-frequency fluctuations of the core and outer core plasma play active, but different, roles in the spontaneous L-H transition. (author)

Evaluation of a thermal SCWR core with sub-channel analysis

International Nuclear Information System (INIS)

Liu Xiaojing; Cheng Xu

2008-01-01

A previous study shows that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behaviour than the existing one-row fuel assemblies. With this new developed two-row fuel assembly, a thermal SCWR core design is proposed Assessment of this design is carried out in this paper. The performance of this new core design is investigated with 3-D coupled thermal-hydraulic/neutronic calculations. During the coupling procedure, the thermal-hydraulic behaviour is analyzed using a single-channel code and the neutron-physical performance is computed with a 3-D reactor physical code. This paper presents the main results achieved so far related to the distribution of some neutronic and thermal-hydraulic parameters. Since the power distribution in some fuel assemblies is extremely uneven, sub-channel analysis is applied to the hottest and most non-uniform assembly in the core. The sub-channel analysis is performed with the power and thermal hydraulic parameters from the coupling results. It provides the hot channel factor and the maximal cladding surface temperature more precisely. The power and mass flux distribution in these assemblies are illustrated in detail for the demonstration purpose. The difference of the results evaluated with two different methods, i.e. sub-channel analysis and single-channel analysis, shows the importance of applying sub-channel analysis. A sensitivity analysis of some important parameters is also carried out. (author)

On-line core monitoring with CORE MASTER / PRESTO

International Nuclear Information System (INIS)

Lindahl, S.O.; Borresen, S.; Ovrum, S.

1986-01-01

Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

Development of digital safety system logic and control

International Nuclear Information System (INIS)

Nishikawa, H.; Sakamoto, H.

1995-01-01

Advanced-BWR (ABWR) uses total digital control and instrumentation (C and I) system. In particular, ABWR adopts a newly developed safety system using advanced digital technology. In the presentation the digital safety system design, manufacturing and factory validation test method are shortly overviewed. The digital safety system consists of micro-processor based digital controllers, data and information transmission by optical fibers and human-machine interface using color flat displays. This new developed safety system meet the nuclear safety requirements such as high reliability, independence of divisions, operability and maintainability. (2 refs., 4 figs., 1 tab.)

Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

International Nuclear Information System (INIS)

Nickel, H.

1985-08-01

Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

PhysDoc: A Distributed Network of Physics Institutions: Collecting, Indexing, and Searching High Quality Documents by Using Harvest; The Dublin Core Metadata Initiative: Mission, Current Activities, and Future Directions; Information Services for Higher Education: A New Competitive Space; Intellectual Property Conservancies.

Science.gov (United States)

Severiens, Thomas; Hohlfeld, Michael; Zimmermann, Kerstin; Hilf, Eberhard R.; von Ossietzky, Carl; Weibel, Stuart L.; Koch, Traugott; Hughes, Carol Ann; Bearman, David

2000-01-01

Includes four articles that discuss a variety to topics, including a distributed network of physics institutions documents called PhysDocs which harvests information from the local Web-servers of professional physics institutions; the Dublin Core metadata initiative; information services for higher education in a competitive environment; and…

Optimization of burnable poison disposition for in-core fuel assemblies

International Nuclear Information System (INIS)

Zhong Wenfa; Luo Rong; Zhou Quan

1997-09-01

The optimization of the burnable poison disposition in the initial core loading of the 200 MW nuclear heating reactor (NHR-200), is studied. The mass fraction of the burnable poison is used as the control variable with the objective to minimize the power peaking factor. The flexible tolerance method is used to solve the nonlinear programming optimal problem. The optimization method can be used in reactor physics design, and get a new pattern of initial core which is of reference value. (2 refs., 8 figs., 1 tab.)

Evolution of helium stars: a self-consistent determination of the boundary of a helium burning convective core

International Nuclear Information System (INIS)

Savonije, G.J.; Takens, R.J.

1976-01-01

A generalization of the Henyey-scheme is given that introduces the mass of the convective core and the density at the outer edge of the convective core boundary as unknowns which have to be solved simultaneously with the other unknowns. As a result, this boundary is determined in a physically self-consistent way for expanding as well as contracting cores, i.e. during the Henyey iterative cycle; its position becomes consistent with the overall physical structure of the star, including the run of the chemical abundances throughout the star. Using this scheme, the evolution of helium stars was followed up to carbon ignition for a number of stellar masses. As compared with some earlier investigations, the calculations show a rather large increase in mass of the convective cores during core helium burning. Evolutionary calculations for a 2M(sun) helium star show that the critical mass for which a helium star ignites carbon non-degenerately lies near 2M(sun). (orig.) [de

The whole-core LEU silicide fuel demonstration in the JMTR

Energy Technology Data Exchange (ETDEWEB)

Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1997-08-01

The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.