An experimental study of the flow characteristics of fluidic device in a passive safety injection tank

International Nuclear Information System (INIS)

Cho, Seok; Song, Chul Hwa; Won, Suon Yeon; Min, Kyong Ho; Chung, Moon Ki

1998-01-01

It is considered to adopt passive safety injection tank (SIT) as a enhanced safety feature in KNGR. Passive SIT employs a vortex chamber as a fluidic device, which control injection flow rate passively by the variation of flow resistance produced by vortex intensity within the vortex chamber. To investigate the flow characteristics of the vortex chamber many tests have been carried out by using small-scale test facility. In this report the effects of geometric parameters of vortex chamber on discharge flow characteristics and the velocity measurement result of flow field, measured by PIV, are presented and discussed. (author). 25 refs., 11 tabs., 31 figs

A Simple Fully Passive Safety Option for SMART SBLOCA

International Nuclear Information System (INIS)

Lee, Won Jae

2012-01-01

SMART reactor, an integral pressurized water reactor (iPWR), is developed by KAERI and now under standard design licensing review. Integral reactor design of the SMART has small diameter penetrations below 2 inches at upper parts of reactor pressure vessel (RPV) and the core is located at very lower part. Amount of reactor coolant inventory is around 0.55tons/MWth during normal operations, which is seven times more than that of conventional PWRs. Such intrinsic safety features of the SMART can provide prolonged core cooling during a small-break loss-of-coolant accident (SBLOCA). As an engineered safety feature for SBLOCA, electrically two-train and mechanically four-train active safety injection (SI) systems are provided to refill the RPV, whose safety been proven through safety analysis and experiments. In addition, four-train passive residual heat removal systems (PRHRSs) are provided to remove core decay heat by natural circulation in the secondary side of steam generators during transient and accident conditions. After Fukushima disaster, a passive safety of nuclear power plants has become more emphasized than conventional active safety, even though there are still debates whether it can really insure the realistic safety. Passive safety is defined such that the core safety is ensured for 72 hours after accidents without any active safety systems and operator actions. In light of this, a simple fully passive safety option for SBLOCA is proposed: low-pressure safety injection tanks (SITs) and heat pipes submerged in the PRHRS emergency coolant tanks (ECTs). Post-LOCA long-term cooling after 72 hours is provided by sump recirculation using shutdown cooling system. Realistic analysis method using MARS3.1 is used to derive fully passive safety option, and then to screen design and operating parameters and to demonstrate the safety performance of SITs. SI line break is selected as a reference SBLOCA scenario

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

Availability analysis of the AP600 passive core cooling system

Energy Technology Data Exchange (ETDEWEB)

Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

1996-12-01

The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

Energy Technology Data Exchange (ETDEWEB)

Tuunanen, J. [VTT Energy, Espoo (Finland). Nuclear Energy; Vihavainen, J. [Lappeenranta Univ. of Technology (Finland); D' Auria, F. [Univ. of Pisa (Italy); Kimber, G. [AEA Technology (United Kingdom)

1999-07-01

The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

International Nuclear Information System (INIS)

Tuunanen, J.; D'Auria, F.; Kimber, G.

1999-01-01

The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

Study on diverse passive decay heat removal approach

International Nuclear Information System (INIS)

Lin Qian; Si Shengyi

2012-01-01

One of the most important principles for nuclear safety is the decay heat removal in accidents. Passive decay heat removal systems are extremely helpful to enhance the safety. In currently design of many advanced nuclear reactors, kinds of passive systems are proposed or developed, such as the passive residual heat removal system, passive injection system, passive containment cooling system. These systems provide entire passive heat removal paths from core to ultimate heat sink. Various kinds of passive systems for decay heat removal are summarized; their common features or differences on heat removal paths and design principle are analyzed. It is found that, these passive decay heat removal paths are similarly common on and connected by several basic heat transfer modes and steps. By the combinations or connections of basic modes and steps, new passive decay heat removal approach or diverse system can be proposed. (authors)

A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

International Nuclear Information System (INIS)

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

1995-01-01

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

Analysis of effect of late water injection on RCS repressurization

International Nuclear Information System (INIS)

Tao Jun; Cao Xuewu

2011-01-01

Effect of late water injection on RCS repressurization during high pressure severe accident sequence in a typical PWR was analyzed. As the results shown, late water injection could increase RCS pressure when RPV failed without RCS passive depressurization. Especially in the condition of opening one PORV, RCS pressure could reach high pressure limit when RPV failed and the risk of HPME and DCH was dramatically increased. Integrity of containment could be threatened. However, in the condition of RCS passive depressurization induced by pressurizer surge line creep failure, RCS pressure could be decreased to very low level even only one PORV was opened and two trains of emergency core cooling were implemented. The risk of HPME and DCH was eliminated. The more PORVs were opened, the faster accident progression was and the earlier RPV failed. RCS pressure was a little higher when PRV failed if two trains of emergency core cooling was implemented comparing with the condition with only one train of emergency core cooling. However the time of RPV failure was obviously delayed. From the point of delaying RPV failure and preventing containment early failure of view, the optimized late water injection was opening three PORVs and implementing two trains of emergency core cooling. (authors)

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

Study on diverse passive decay heat removal approach and principle

International Nuclear Information System (INIS)

Lin Qian; Si Shengyi

2012-01-01

Decay heat removal in post-accident is one of the most important aspects concerned in the reactor safety analysis. Passive decay heat removal approach is used to enhance nuclear safety. In advanced reactors, decay heat is removed by multiple passive heat removal paths through core to ultimate heat sink by passive residual heat removal system, passive injection system, passive containment cooling system and so on. Various passive decay heat removal approaches are summarized in this paper, the common features and differences of their heat removal paths are analyzed, and the design principle of passive systems for decay heat removal is discussed. It is found that. these decay heat removal paths is combined by some basic heat transfer processes, by the combination of these basic processes, diverse passive decay heat removal approach or system design scheme can be drawn. (authors)

Evaluation of the gravity-injection capability for core cooling after a loss-of-SDC event

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

1999-01-01

In order to evaluate the gravity-drain capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Young Gwang Units 3 and 4 were reviewed. The six cases of possible gravity-drain paths using the water of the refueling water storage tank (RWST) were identified and the thermal hydraulic analyses were performed using RELAP5/MOD3.2 code. The core cooling capability was dependent on the gravity-drain paths and the drain rate. In the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. However, in the cases with the injection path and opening on the cold leg side, the core coolant continued boiling although the system pressure remains atmospheric after gravity-injection because the cold water injected from the RWST was bypassed the core region. In the cases with the higher pressurizer opening than the RWST water level, the system was also pressurized by the water-hold in the pressurizer and the core was uncovered because the gravity-injection from the RWST stopped due to the high system pressure. In addition, from the sensitivity study on the gravity-injection flow rates, it was found that about 54 kg/s of RWST drain rate was required to maintain the core cooling. Those analysis results would provide useful information to operators coping with the event

Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

International Nuclear Information System (INIS)

Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

1995-01-01

The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

Proposal for a advanced PWR core with adequate characteristics for passive safety concept

International Nuclear Information System (INIS)

Perrotta, Jose Augusto

1999-01-01

This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

Analysis of emergency core cooling capability of direct vessel vertical injection using CFX

International Nuclear Information System (INIS)

Yoon, Sang H.; Yu, Yong H.; Suh, Kune Y.

2003-01-01

More reliable and efficient safety injection system is of utmost importance in the design of advanced reactors such as the APR1400 (Advanced Power Reactor 1400 MWe). In this work, a new idea is proposed to inject the Emergency Core Cooling (ECC) water utilizing a dedicated nozzle with a vertically downward elbow. The Direct Vessel Injection (DVI) system is located horizontally above the cold leg in the APR1400. However, the horizontal injection method may not always satisfy the ECC penetration requirement into the core on account of rather involved multidimensional thermal and hydraulic phenomena occurring in the annular reactor downcomer such as bypass, impingement, entrainment and sweepout, condensation oscillation, etc. Thus, a novel concept is called for from the reactor safety point of view. The Direct Vessel Vertical Injection (DVVI) system is one of these efforts to penetrate as much the ECC water through the downcomer into the core as is practically achievable. The DVVI system can increase the momentum of the downward flow, thus minimizing the effect of water impingement on the core barrel and the direct bypass though the break. To support the claim of increased downward momentum of flow in the DVVI system, computational fluid dynamics analyses were performed using CFX. The new concept of the DVVI system, which can certainly help increase the core thermal margin, is found to be more efficient than DVI. If the structural problem in the manufacturing process is properly solved, this concept can safely be applied in the advanced nuclear reactor design

No change in calf muscle passive stiffness after botulinum toxin injection in children with cerebral palsy.

Science.gov (United States)

Alhusaini, Adel A A; Crosbie, Jack; Shepherd, Roberta B; Dean, Catherine M; Scheinberg, Adam

2011-06-01

Stiffness and shortening of the calf muscle due to neural or mechanical factors can profoundly affect motor function. The aim of this study was to investigate non-neurally mediated calf-muscle tightness in children with cerebral palsy (CP) before and after botulinum toxin type A (BoNT-A) injection. Sixteen children with spastic CP (seven females, nine males; eight at Gross Motor Function Classification System level I, eight at level II; age range 4-10 y) and calf muscle spasticity were tested before and during the pharmaceutically active phase after injection of BoNT-A. Measures of passive muscle compliance and viscoelastic responses, hysteresis, and the gradient of the torque-angle curve were computed and compared before and after injection. Although there was a slight, but significant increase in ankle range of motion after BoNT-A injection and a small, significant decrease in the torque required to achieve plantigrade and 5° of dorsiflexion, no significant difference in myotendinous stiffness or hysteresis were detected after BoNT-A injection. Despite any effect on neurally mediated responses, the compliance of the calf muscle was not changed and the muscle continued to offer significant resistance to passive motion of the ankle. These findings suggest that additional treatment approaches are required to supplement the effects of BoNT-A injections when managing children with calf muscle spasticity. © The Authors. Developmental Medicine & Child Neurology © 2011 Mac Keith Press.

The effects of intrahippocampus injection of progesterone on passive avoidance learning and memory in adult male rats

Directory of Open Access Journals (Sweden)

Tahmoores Shahrivar

2014-10-01

Full Text Available Background: Progesterone as a reproductive hormone is studied in the brain structure. New studies are mainly about systemic injection and have shown that the progesterone has different effects on memory. But it is not clear if hippocampus is mediator of progesterone in reinforcement of memory and learning. So, the effects of intra-hippocampal injection of progesterone on passive avoidance memory and learning was studied in male Wistar rats. Material and Methods: Animals were divided in eightgroups as a control, saline, almond oil (as a vehicle and progesterone groups (0.5, 1, 1.5, 2, 3µg/Kg. body weight. All groups except the control were cannulated in left CA1 area by stereotaxic device. One week after recovery, all groups were trained by shuttle box and immediately after training different doses of progesterone in the volume of 0.5μL were injected through the cannula in CA1 of hippocampus and control groups received saline and almond oil at the same volume and condition respectively. After 48 hours passive avoidance memory and learning was evaluated. Results: Statistical analysis showed that 0.5, 1.5, 2 µg/µl progesterone groups had significantly improved passive avoidance learning and memory compared with saline group (p <0.001, but 1 and 3 µg/µl progesterone groups did not show significant effect on passive avoidance learning and memory compared with saline group. Also almond oil increased passive avoidance learning and memory compared with saline group (p <0.001. Conclusion: It seems that effects of progesterone as a neurosteroid on memory and learning depend on its dose. Also almond oil contains vitamin E, omega 3 and oleic acid that has beneficial effects on learning and memory.

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Zhang-cheng; Xie, Heng

2014-01-01

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

Comparative Investigation on 0.4 inch SBLOCA Scenario with Single and Dual Train Passive Safety Injection Systems using SMART-ITL

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yun, Eun-Koo; Choi, Nam-Hyun; Min, Kyoung-Ho; Shin, Yong-Cheol; Bang, Yoon-Gon; Kim, Myoung-Jun; Seo, Chan-Jong; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The Standard Design Approval (SDA) for SMART was certificated in 2012 at the Korea Atomic Energy Research Institute (KAERI). In December 2015, Saudi Arabia and Korea started conducting a three-year project of Pre-Project Engineering (PPE) to prepare a Preliminary Safety Analysis Report (PSAR) and to review the feasibility of constructing SMART reactors in Saudi Arabia. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMART-ITL. In this paper, a comparative investigation was performed on 0.4 inch SBLOCA scenario with single and dual train passive safety injection systems using SMART-ITL. In this paper, the effect of the train number of PSIS on a SBLOCA scenario is investigated for a break size of 0.4 inch. The single and dual train tests show a similar trend in general but the injected water migrates slightly differently in the RV and is discharged through the break nozzle. The parameters of the RV pressure, RV water level, accumulated break mass, and injection flowrates from the CMT and SIT were compared. Compared with the single train test, the increased injection rates from the two trains of the PSIS during the dual train test raised the RV water level, ensuring the safety of the reactor core.

A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

International Nuclear Information System (INIS)

Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

2015-01-01

Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

2015-12-15

Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

International Nuclear Information System (INIS)

Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

2001-01-01

Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

Concepts for passive heat removal and filtration systems under core meltdown conditions

International Nuclear Information System (INIS)

Wilhelm, J.G.; Neitzel, H.-J.

1993-01-01

The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

International Nuclear Information System (INIS)

Yoon, Sang H.; Lee, Jong G.; Suh, Kune Y.

2006-01-01

Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400

Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

International Nuclear Information System (INIS)

Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

1979-01-01

Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

Gravity driven emergency core cooling experiments with the PACTEL facility

International Nuclear Information System (INIS)

Munther, R.; Kalli, H.; Kouhia, J.

1996-01-01

PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs

Gravity driven emergency core cooling experiments with the PACTEL facility

Energy Technology Data Exchange (ETDEWEB)

Munther, R; Kalli, H [University of Technology, Lappeenranta (Finland); Kouhia, J [Technical Research Centre of Finland, Lappeenranta (Finland)

1996-12-01

PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs.

CANDU passive shutdown systems

Energy Technology Data Exchange (ETDEWEB)

Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

1996-12-01

CANDU incorporates two diverse, passive shutdown systems, independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

LOCA analysis of SCWR-M with passive safety system

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

2013-06-15

Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

Design and development of innovative passive valves for Nuclear Power Plant applications

Energy Technology Data Exchange (ETDEWEB)

Sapra, M.K., E-mail: sapramk@barc.gov.in; Kundu, S.; Pal, A.K.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

2015-05-15

by India. For example, the Hot Shutdown Passive Valves (HSPV), developed for the decay heat removal system keep the main heat transport system under hot conditions by passively sensing and controlling the system pressure. Another crucial and important valve which has been successfully developed is the Poison Injection Passive Valve (PIPV) for the Passive Poison Injection System. It not only provides higher reliability, but also ensures safe shutdown of the reactor in case of insider threats or malevolent acts in disabling active shutdown system of the reactor. Recently, an innovative valve called the Accumulator Isolation Passive Valve (AIPV) has been developed for the Emergency Core Cooling System (ECCS), which is engineered to mitigate the consequences of Loss of Coolant Accident (LOCA). During normal operation of the reactor, the pressurized accumulators (55 bar) are kept isolated from the reactor core (70 bar) by means of AIPVs. In case of a LOCA, these passive valves open when the main heat transport system pressure falls to a desired value. For prolonged cooling of the core, these passive valves regulate the discharge in a desired manner. These are non-standard, high pressure and high temperature valves, which are unavailable commercially and hence have to be indigenously designed and developed. This paper primarily deals with the design, development and testing of Accumulator Isolation Passive Valves (AIPV) proposed to be used in the ECCS. A 25 NB size AIPV has been designed and successfully tested at Integral Test Loop (ITL) under simulated reactor conditions. It is a self-acting, ANSI 600 rating valve, which requires no external energy (i.e., neither air nor electrical power). It not only provides passive isolation but also passively controls high pressure liquid discharge through it. The design concept of the valve, functional performance, in situ valve testing methodology and the test results at simulated conditions are discussed.

Safety significance of ATR passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1990-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety of the facility. The three passive safety attributes being evaluated in the paper are: 1) In-core and in-vessel natural convection cooling, 2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and 3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond to most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) models and results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR firewater injection system (emergency coolant system)

Single-mode Brillouin fiber laser passively stabilized at resonance frequency with self-injection locked pump laser

International Nuclear Information System (INIS)

Spirin, V V; Lopez-Mercado, C A; Megret, P; Fotiadi, A A

2012-01-01

We demonstrate a single-mode Brillouin fiber ring laser, which is passively stabilized at pump resonance frequency by using self-injection locking of semiconductor pump laser. Resonance condition for Stokes radiation is achieved by length fitting of Brillouin laser cavity. The laser generate single-frequency Stokes wave with linewidth less than 0.5 kHz using approximately 17-m length cavity

Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

International Nuclear Information System (INIS)

Morimoto, Y.; Madarame, H.; Okamoto, K.

2001-01-01

Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

International Nuclear Information System (INIS)

Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young

2013-01-01

Highlights: • We newly propose the design concept of integrated passive safety system (IPSS). • It has five safety functions for decay heat removal and severe accident mitigation. • Simulations for IPSS show that core melt does not occur in accidents with SBO. • IPSS can achieve the passive in-vessel retention and ex-vessel cooling strategy. • The applicability of IPSS is high due to the installation outside the containment. -- Abstract: The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1985-07-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Study on core make-up water experiment of AC600 make-up water tank

International Nuclear Information System (INIS)

Ji Fuyun; Li Changlin; Zheng Hua; Liu Shaohua; Xu Xiaolan

1999-01-01

The core makeup tank (CMT) is a principal component of the passive high pressure safety injection systems for AC600 and has a function to inject cold borated water into reactor vessel during abnormal events. The purpose of this experiment is to verify the gravity drain behavior of the CMT and to provide experimental data to verify the computer codes used in the safety analyses. Five experiments with simulative small and medium break conditions are conducted at AC600 core makeup tank performance test facility of Nuclear Power Institute of China (NPIC). The author provides the results of one test. The simulated accident is a small break loss-of-coolant accident

PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

International Nuclear Information System (INIS)

1991-03-01

An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

Energy Technology Data Exchange (ETDEWEB)

Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

An experimental study on passive safety systems for the SMART design with the SMART-ITL facility

International Nuclear Information System (INIS)

Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yi, Sung-Jae

2016-01-01

Passive Safety Systems (PSSs) are added to the SMART design to increase the safety margin during accidents especially under a prolonged station blackout. A set of validation tests were performed for the PSSs of the SMART design with an integral effect test loop of SMART-ITL. Both single and dual trains of the Passive Safety Injection System (PSIS) were simulated to validate the SMART design together with two stages of Automatic Depressurization System (ADS) and four trains of Passive Residual Heat Removal System (PRHRS), and their results were compared. In this paper, the effect of the train number of PSIS on a Small-Break Loss of Coolant Accident (SBLOCA) scenario is investigated for a break size of 0.4 inch. The single and dual train tests show a similar trend in general but the injected water migrates slightly differently in the RV and is discharged through the break nozzle. The parameters of the Reactor Vessel (RV) pressure, RV water level, accumulated break mass, and injection flowrates from the Core Makeup Tank (CMT) and Safety Injection Tank (SIT) were compared. The acquired data will be used to validate the safety analysis code and its related models to evaluate the performance of SMART PSS, and to provide the base data during the application phase of construction licensing of the SMART design. (author)

Application of a steam injector for passive emergency core cooling during a station blackout

International Nuclear Information System (INIS)

Heinze, D.; Behnke, L.; Schulenberg, T.

2012-01-01

One of the basic protection targets of reactor safety is the safe heat removal during normal operation but also following shut-down. Since the reactor accident in Fukushima an optimization of the plant robustness in case of beyond-design accident is performed. Special attention is given to the increase of time available for starting appropriate measures for emergency core cooling in case of a station blackout. The state-of the art in engineering and research is presented. Investigations on the applicability of a steam injector for passive emergency core cooling during a station blackout in BWR-type reactors have progressed, experiments on dynamic behavior of the injector are described. A precise design with respect to the thermal hydraulic boundary conditions has been performed.

Transient simulation of ALWR passive safety systems using RELAP5/MOD2

International Nuclear Information System (INIS)

Elias, E.; Nekhamkin, Y.; Arshavski, I.

2004-01-01

Numerical simulation is presented of some passive safety systems currently incorporated in the design of the next generation advanced light water reactors (ALWRs). The performance and effectiveness of ex-core natural convection cooling and the concept of gravity driven water injection at high pressure are investigated using the RELAP5/MOD2 thermal-hydraulic code. The study identifies areas that should be investigated more fully in future experimental programs related to hypothetical large and small LOCA in ALWRs. (author)

Technical feasibility and reliability of passive safety systems of AC600

International Nuclear Information System (INIS)

Niu, W.; Zeng, X.

1996-01-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished by the Nuclear Power Institute of China. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also

Technical feasibility and reliability of passive safety systems of AC600

Energy Technology Data Exchange (ETDEWEB)

Niu, W; Zeng, X [Nuclear Power Inst. of China, Chendu (China)

1996-12-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also involved. (author). 3 figs, 1 tab.

TRIGGERING COLLAPSE OF THE PRESOLAR DENSE CLOUD CORE AND INJECTING SHORT-LIVED RADIOISOTOPES WITH A SHOCK WAVE. II. VARIED SHOCK WAVE AND CLOUD CORE PARAMETERS

Energy Technology Data Exchange (ETDEWEB)

Boss, Alan P.; Keiser, Sandra A., E-mail: boss@dtm.ciw.edu, E-mail: keiser@dtm.ciw.edu [Department of Terrestrial Magnetism, Carnegie Institution, 5241 Broad Branch Road, NW, Washington, DC 20015-1305 (United States)

2013-06-10

A variety of stellar sources have been proposed for the origin of the short-lived radioisotopes that existed at the time of the formation of the earliest solar system solids, including Type II supernovae (SNe), asymptotic giant branch (AGB) and super-AGB stars, and Wolf-Rayet star winds. Our previous adaptive mesh hydrodynamics models with the FLASH2.5 code have shown which combinations of shock wave parameters are able to simultaneously trigger the gravitational collapse of a target dense cloud core and inject significant amounts of shock wave gas and dust, showing that thin SN shocks may be uniquely suited for the task. However, recent meteoritical studies have weakened the case for a direct SN injection to the presolar cloud, motivating us to re-examine a wider range of shock wave and cloud core parameters, including rotation, in order to better estimate the injection efficiencies for a variety of stellar sources. We find that SN shocks remain as the most promising stellar source, though planetary nebulae resulting from AGB star evolution cannot be conclusively ruled out. Wolf-Rayet (WR) star winds, however, are likely to lead to cloud core shredding, rather than to collapse. Injection efficiencies can be increased when the cloud is rotating about an axis aligned with the direction of the shock wave, by as much as a factor of {approx}10. The amount of gas and dust accreted from the post-shock wind can exceed that injected from the shock wave, with implications for the isotopic abundances expected for a SN source.

Hydro-geophysical responses to the injection of CO2 in core plugs of Berea sandstone

Science.gov (United States)

Song, I.; Park, K. G.

2017-12-01

We have built a laboratory-scale core flooding system to measure the relative permeability of a core sample and the acoustic response to the CO2 saturation degree at in situ condition of pressure and temperature down to a few kilometer depths. The system consisted of an acoustic velocity core holder (AVC model from the Core Laboratories) between upstream where CO2 and H2O were injected separately and downstream where the mixed fluids came out of a core sample. Core samples with 4 cm in diameter and 5 cm in length of Berea sandstone were in turn placed in the core holder for confining and axial pressures. The flooding operations of the multiphase fluids were conducted through the sample at 40ºC in temperature and 8 MPa in backpressure. CO2 and H2O in the physical condition were injected separately into a sample at constant rate with various ratios. The two phases were mixed during flowing through the sample. The mixed fluids out of the sample were separated again by their different densities in a chamber equipped with a level gauge of the interface. From the level change of the water in the separator, we measured the volume of water coming out of the sample for each test with a constant ratio of the injection rates. Then it was possible to calculate the saturation degree of CO2 from the difference between input volume and output volume of water. The differential pressure between upstream and downstream was directly measured to calculate the relative permeability as a function of the CO2 saturation degree. We also conducted ultrasonic measurements using piezoelectric sensors on the end plugs. An electric pulse was given to a sensor on one end of sample, and then ultrasonic waves were recorded from the other end. The various ratios of injection rate of CO2 and H2O into Berea sandstone yielded a range of 0.1-0.7 in CO2 saturation degree. The relative permeability was obtained at the condition of steady-state flow for given stages from the velocity of each phase and

Linear stability analysis of the gas injection augmented natural circulation of STAR-LM

International Nuclear Information System (INIS)

Yeon-Jong Yoo; Qiao Wu; James J Sienicki

2005-01-01

Full text of publication follows: A linear stability analysis has been performed for the gas injection augmented natural circulation of the Secure Transportable Autonomous Reactor - Liquid Metal (STAR-LM). Natural circulation is of great interest for the development of Generation-IV nuclear energy systems due to its vital role in the area of passive safety and reliability. One of such systems is STAR-LM under development by Argonne National Laboratory. STAR-LM is a 400 MWt class modular, proliferation-resistant, and passively safe liquid metal-cooled fast reactor system that uses inert lead (Pb) coolant and the advanced power conversion system that consists of a gas turbine Brayton cycle utilizing supercritical carbon dioxide (CO 2 ) to obtain higher plant efficiency. The primary loop of STAR-LM relies only on the natural circulation to eliminate the use of circulation pumps for passive safety consideration. To enhance the natural circulation of the primary coolant, STAR-LM optionally incorporates the additional driving force provided by the injection of noncondensable gas into the primary coolant above the reactor core, which is effective in removing heat from the core and transferring it to the secondary working fluid without the attainment of excessive coolant temperature at nominal operating power. Therefore, it naturally raises the concern about the natural circulation instability due to the relatively high temperature change in the core and the two-phase flow condition in the hot leg above the core. For the ease of analysis, the flow path of the loop was partitioned into five thermal-hydraulically distinct sections, i.e., heated core, unheated core, hot leg, heat exchanger, and cold leg. The one-dimensional single-phase flow field equations governing the natural circulation, i.e., continuity, momentum, and energy equations, were used for each section except the hot leg. For the hot leg, the one-dimensional homogeneous equilibrium two-phase flow field

Passive safety system of a super fast reactor

Energy Technology Data Exchange (ETDEWEB)

Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

2015-08-15

Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

International Nuclear Information System (INIS)

Lioce, D.; Asztalos, M.; Alemberti, A.; Barucca, L.; Frogheri, M.; Saiu, G.

2012-01-01

Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000 ® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

Energy Technology Data Exchange (ETDEWEB)

Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1989-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

Passive systems for light water reactors

International Nuclear Information System (INIS)

Adinolfi, R.; Noviello, L.

1990-01-01

The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

Thermal-hydraulic modeling needs for passive reactors

Energy Technology Data Exchange (ETDEWEB)

Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

Thermal-hydraulic modeling needs for passive reactors

International Nuclear Information System (INIS)

Kelly, J.M.

1997-01-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken

Successful Treatment of Early Talar Osteonecrosis by Core Decompression Combined with Intraosseous Stem Cell Injection: A Case Report.

Science.gov (United States)

Nevalainen, Mika T; Repo, Jussi P; Pesola, Maija; Nyrhinen, Jukka P

2018-01-01

Osteonecrosis of the talus is a fairly rare condition. Many predisposing factors have been identified including previous trauma, use of corticosteroids, alcoholism, and smoking. As a gold standard, magnetic resonance imaging (MRI) is the most sensitive and specific diagnostic examination to detect osteonecrosis. While many treatment options for talar osteonecrosis exist, core decompression is suggested on young patients with good outcome results. More recently, intraosseous stem cell and platelet-rich plasma (PRP) injection has been added to the core decompression procedure. We report a successful treatment of early talar osteonecrosis ARCO I (Association Research Circulation Osseous) by core decompression combined with stem cell and PRP injection. On 3-month and 15-month follow-up, MRI showed complete resolution of the osteonecrotic changes together with clinical improvement. This modified technique is a viable treatment option for early talar osteonecrosis. Nevertheless, future prospects should include a study comparing this combined technique with plain core decompression.

A fast alternative to core plug tests for optimising injection water salinity for EOR

DEFF Research Database (Denmark)

Hassenkam, Tue; Andersson, Martin Peter; Hilner, Emelie Kristin Margareta

2014-01-01

of the clays which would lead to permanent reservoir damage but evidence of effectiveness at moderate salinity would offer the opportunity to dispose of produced water. The goal is to define boundary conditions so injection water salinity is high enough to prevent reservoir damage and low enough to induce...... the low salinity effect while keeping costs and operational requirements at a minimum. Traditional core plug testing for optimising conditions has some limitations. Each test requires a fresh sample, core testing requires sophisticated and expensive equipment, and reliable core test data requires several...... experiments can be done relatively quickly on very little material, it gives the possibility of testing salinity response on samples from throughout a reservoir and for gathering statistics. Our approach provides a range of data that can be used to screen core plug testing conditions and to provide extra data...

Combined active-passive immunisation of horses against tetanus.

Science.gov (United States)

Liefman, C E

1980-03-01

The protection afforded by active, passive and combined active-passive methods of immunisation against tetanus was examined in previously unimmunised horses. Three groups of horses were injected; one with tetanus toxoid alone, one with tetanus antitoxin alone and one in which the tetanus toxoid and tetanus antitoxin were injected simultaneously. The protection afforded was determined by monitoring the levels of antitoxin achieved in the horses by each of these methods. The results obtained demonstrated the effectiveness of the combined active-passive method in affording rapid and prolonged protection and enabled the examination of some of the factors involved in active and in passive immunisation when used alone. The advantages obtained by the use of the combined active-passive method in protecting unimmunised horses suddenly placed at risk to infection are outlined.

Fabrication and modelling of injection moulded all-polymer capillary microvalves for passive microfluidic control

International Nuclear Information System (INIS)

Kistrup, Kasper; Østergaard, Peter Friis; Taboryski, Rafael; Wolff, Anders; Hansen, Mikkel Fougt; Haugshøj, Kenneth Brian; Poulsen, Carl Esben

2014-01-01

Rapid prototyping is desirable when developing products. One example of such a product is all-polymer, passive flow controlled lab-on-a-chip systems that are preferential when developing low-cost disposable chips for point-of-care use. In this paper we investigate the following aspects of going from rapid prototyping to pilot (mass) production. (1) Fabrication of an all-polymer microfluidic system using a rapid prototyped master insert for injection moulding and ultrasonic welding, including a systematic experimental characterisation of chip featured geometric capillary microvalve test structures. (2) Numerical modelling of the microvalve burst pressures. Numerical modelling of burst pressures is challenging due to its non-equilibrium nature. We have implemented and tested the level-set method modified with a damped driving term and show that the introduction of the damping term leads to numerically robust results with limited computational demands and a low number of iterations. Numerical and simplified analytical results are validated against the experimental results. We find that injection moulding and ultrasonic welding are effective for chip production and that the experimental burst pressures could be estimated with an average accuracy of 5% using the presented numerical model. (paper)

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

International Nuclear Information System (INIS)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

2014-01-01

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Laboratory Mid-frequency (Kilohertz) Range Seismic Property Measurements and X-ray CT Imaging of Fractured Sandstone Cores During Supercritical CO2 Injection

Science.gov (United States)

Nakagawa, S.; Kneafsey, T. J.; Chang, C.; Harper, E.

2014-12-01

During geological sequestration of CO2, fractures are expected to play a critical role in controlling the migration of the injected fluid in reservoir rock. To detect the invasion of supercritical (sc-) CO2 and to determine its saturation, velocity and attenuation of seismic waves can be monitored. When both fractures and matrix porosity connected to the fractures are present, wave-induced dynamic poroelastic interactions between these two different types of rock porosity—high-permeability, high-compliance fractures and low-permeability, low-compliance matrix porosity—result in complex velocity and attenuation changes of compressional waves as scCO2 invades the rock. We conducted core-scale laboratory scCO2 injection experiments on small (diameter 1.5 inches, length 3.5-4 inches), medium-porosity/permeability (porosity 15%, matrix permeability 35 md) sandstone cores. During the injection, the compressional and shear (torsion) wave velocities and attenuations of the entire core were determined using our Split Hopkinson Resonant Bar (short-core resonant bar) technique in the frequency range of 1-2 kHz, and the distribution and saturation of the scCO2 determined via X-ray CT imaging using a medical CT scanner. A series of tests were conducted on (1) intact rock cores, (2) a core containing a mated, core-parallel fracture, (3) a core containing a sheared core-parallel fracture, and (4) a core containing a sheared, core-normal fracture. For intact cores and a core containing a mated sheared fracture, injections of scCO2 into an initially water-saturated sample resulted in large and continuous decreases in the compressional velocity as well as temporary increases in the attenuation. For a sheared core-parallel fracture, large attenuation was also observed, but almost no changes in the velocity occurred. In contrast, a sample containing a core-normal fracture exhibited complex behavior of compressional wave attenuation: the attenuation peaked as the leading edge of

Effect of capillary number on the oil recovery using oil-water emulsion injection in core flooding experiments

Energy Technology Data Exchange (ETDEWEB)

Guillen Nunez, Victor Raul; Carvalho, Marcio da Silveira [Pontifical Catholic University of Rio de Janeiro (PUC-Rio), RJ (Brazil). Dept. of Mechanical Engineering], E-mail: msn@puc-rio.br; Basante, Vladimir Alvarado [University of Wyoming, Laramie, WY (United States). Dept. of Chemical/Petroleum Engineering], E-mail: valvard@uwyo.edu

2010-07-01

The Water injection flooding is a common method to improve reservoir sweep and pressure maintenance. The heavy-oil-recovery efficiency is in part limited by the high water-to-oil mobility ratio. Several enhanced oil recovery methods are being developed as more efficient alternatives to water flooding. Dispersion injection, in particular oil-water emulsion injection, has been tried with relative success as an enhanced oil recovery method, but the technique is not fully developed or understood. If emulsion injection proves to be an effective EOR method, its use would bring the added benefit of disposing produced water with small oil content that could be modified to serve as the injected oil-water emulsion. The use of such methods requires a detailed analysis of the different flow regimes of emulsions through the porous space of a reservoir rock. If the drop size of the disperse phase is of the same order of magnitude as the pore size, the drops may agglomerate and partially block water flow through pores. This flow regime may be used to control the mobility of the injected liquid, leading to higher recovery factor. We have shown in recent experiments of oil displacement in a sandstone core that, the oil recovery factor could be raised from approximately 40 %, obtained with water injection only, up to approximately 75 % by alternating water and emulsion injection. Although these results clearly show the improvement in the recovery factor, the mechanisms responsible for the phenomenon have not been clearly elucidated. In this work, two sandstone cores were used to demonstrate the effect of flow rate (capillary number) on the mobility control by emulsion injection. Figure 1 shows a schematic representation of the experiment set-up. The experiments show that raising the flow rate by a factor of 10 (0.03 ml/min to 0.3 ml/min), the oil recovered factor decreases considerably. (author)

Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

Energy Technology Data Exchange (ETDEWEB)

Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-06-15

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

Modern Cored Wire Injection 2PE-9 Method in the Production of Ductile Iron

Directory of Open Access Journals (Sweden)

E. Guzik

2012-04-01

Full Text Available The results of studies on the use of modern two cored wires injection method for production of nodular graphite cast iron with use of unique implementation of drum ladle as a treatment/ transport and casting ladle instead vertical treatment ladle was described. The injection of length of Ø 9mm wires, cored: in FeSi + Mg nodulariser mixture and inoculant master alloy is a treatment method which can be used to produce iron melted in coreless induction furnace. This paper describes the results of using this method for possibility production of ductile iron under specific industrial conditions. In this case was taken ductile iron with material designation: EN-GJS-450- 10 Grade according PN-EN 1563:2000. Microstructure of 28 trials was controlled on internally used sample which has been correlated with standard sample before. The paper presents typical metallic matrix and graphite characteristic. Additionally, mechanical properties were checked in one experiment. Because of further possibility treatment temperature reduction only the rough magnesium recovery and cost of this new method are given.

Performance of ARCHITECT HCV core antigen test with specimens from US plasma donors and injecting drug users.

Science.gov (United States)

Mixson-Hayden, Tonya; Dawson, George J; Teshale, Eyasu; Le, Thao; Cheng, Kevin; Drobeniuc, Jan; Ward, John; Kamili, Saleem

2015-05-01

Hepatitis C virus (HCV) core antigen is a serological marker of current HCV infection. The aim of this study was mainly to evaluate the performance characteristics of the ARCHITECT HCV core antigen assay with specimens from US plasma donors and injecting drug users. A total of 551 serum and plasma samples with known anti-HCV and HCV RNA status were tested for HCV core antigen using the Abbott ARCHITECT HCV core antigen test. HCV core antigen was detectable in 100% of US plasma donor samples collected during the pre-seroconversion phase of infection (anti-HCV negative/HCV RNA positive). Overall sensitivity of the HCV core antigen assay was 88.9-94.3% in samples collected after seroconversion. The correlation between HCV core antigen and HCV RNA titers was 0.959. HCV core antigen testing may be reliably used to identify current HCV infection. Published by Elsevier B.V.

Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

Fabrication and modelling of injection moulded all-polymer capillary microvalves for passive microfluidic control

DEFF Research Database (Denmark)

Kistrup, Kasper; Poulsen, Carl Esben; Østergaard, Peter Friis

2014-01-01

Rapid prototyping is desirable when developing products. One example of such a product is all-polymer, passive flow controlled lab-on-a-chip systems that are preferential when developing low-cost disposable chips for point-of-care use. In this paper we investigate the following aspects of going...... from rapid prototyping to pilot (mass) production. (1) Fabrication of an all-polymer microfluidic system using a rapid prototyped master insert for injection moulding and ultrasonic welding, including a systematic experimental characterisation of chip featured geometric capillary microvalve test...... structures. (2) Numerical modelling of the microvalve burst pressures. Numerical modelling of burst pressures is challenging due to its non-equilibrium nature. We have implemented and tested the level-set method modified with a damped driving term and show that the introduction of the damping term leads...

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

1991-01-01

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

Application of hepatitis B core particles produced by human primary hepatocellular carcinoma (PLC/342) propagated in nude mice to the determination of anti-HBc by passive hemagglutination.

Science.gov (United States)

Miyamoto, K; Itoh, Y; Tsuda, F; Matsui, T; Tanaka, T; Miyamoto, H; Naitoh, S; Imai, M; Usuda, S; Nakamura, T

1986-05-22

Human primary hepatocellular carcinoma (PLC/342), carried by nude mice, produces hepatitis B core particles as well as hepatitis B surface antigen particles. Core particles purified form PLC/342 tumors displayed epitopes of hepatitis B core antigen (HBcAg) but not epitopes of hepatitis B e antigen (HBeAg) on their surface, unlike core particles prepared from Dane particles, derived from plasma of asymptomatic carriers, that expressed epitopes of both HBcAg and HBeAg. Core particles obtained from PLC/342 tumors were applied to the determination of antibody to HBcAg (anti-HBc) by passive hemagglutination. The assay detected anti-HBc not only in individuals with persistent infection with hepatitis B virus and in those who had recovered from transient infection, but also in patients with acute type B hepatitis, indicating that it can detect anti-HBc of either IgG or IgM class. A liberal availability of core particles from tumors carried by nude mice, taken together with an easy applicability of the method, would make the passive hemagglutination for anti-HBc a valuable tool in clinical and epidemiological studies, especially in places where sophisticated methods are not feasible.

Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

2013-10-15

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

Directory of Open Access Journals (Sweden)

VIKAS JAIN

2013-10-01

Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

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.

Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

The Heat Exchanger for Passive Part ECCS of WWER-1000 on Base of the Thermo siphons

International Nuclear Information System (INIS)

Kirov, V.; Chulkin, O.

2008-01-01

One of NPP's systems providing safe operation is the system of emergency core cooling system (ECCS), which primary function in accidents is to flood the nuclear reactor core and to assure the sub critical condition and core cooling. At injection of cold water in reactor thermal stresses and thermal fatigue in the vessel cladding and constructional materials are arise. Low temperature of the water injected in reactor is a reason of occurrence of these undesirable consequences. Some variants of the water heating in accumulators of ECCS are considered. Now at Ukrainian NPPs the electrical heating in accumulators is used. Electrical heaters create the essential additional loading to diesel generators at imposing of two accidents - the large break and losses of power supplies on own needs. It is offered to use a heater in accumulators that working by a principle two-phase thermal siphon which advantages is: small dimensions, small delay and design reliability. In such heat exchanger the heating medium is a direct steam and the heated up medium is water with boric acid from accumulators of ECCS. Under requirements of the service regulations of ECCS accumulators it is necessary to guarantee injected water heating up to 90 ?? in case of a small break and to 150 ?? in case of the large break. Results of calculations for different external diameters of a tube of thermal siphon which have allowed to define the constructive sizes of heat exchanger, providing necessary conditions for required functioning of passive part ECCS are submitted The calculation and analysis of operating modes of the changed circuit of passive part ECCS for various accidents is carried out. The calculated pressure drop indicates that changes do not have essential influence on system work as a whole. Thus, the submitted decision provides the increase of reliability of ECCS at small and large breaks accidents, i.e. in all modes stipulated by the project.(author)

The Effect of Temperature and Injection Rate during Water Flooding Using Carbonate Core Samples: An Experimental Approach

Directory of Open Access Journals (Sweden)

Yaser Ahmadi

2016-10-01

Full Text Available In many reservoirs, after water flooding, a large volume of oil is still left behind. Hot water injection is the most basic type of thermal recovery which increase recovery by improved sweep efficiency and thermal expansion of crude.In the present work, the effects of injection rate and the temperature of the injected water were surveyed by using core flooding apparatus. Water flooding was performed at different rates (0.2, 0.3, and 0.4 cc/min and temperatures (20 and 90 °C, and the reservoir temperature was about 63 °C. Oil recovery during hot water injection was more than water injection. Moreover, it was concluded that at injection rates of 0.2, 0.3, and 0.4 cc/min breakthrough time in hot water injection occurred 10 min later in comparison to water injection. The results showed that higher oil recovery and longer breakthrough time were obtained as a result of reducing injection rate. In the first 50 minutes, the oil recovery at injection rates of 0.2, 0.3 and 0.4 cc/min was 27.5, 34, and 46% respectively. It was found that at the beginning of injection, thermal and non-thermal injection recovery factors are approximately equal. Moreover, according to the results, recovery factor at the lowest rate in hot water (T=90 °C and q=0.2 cc/min is the best condition to obtain the highest recovery.

RELAP5 analysis of PACTEL injection tests

International Nuclear Information System (INIS)

Kimber, G.R.; Lillington, J.N.

2000-01-01

A characteristic feature of advanced reactor designs is their reliance on passive safety systems. It is important to assess both the operation of such systems and the ability of systems codes, such as RELAP5, to model them. In Finland VTT Energy, together with Lappeenranta University of Technology, is using the PACTEL facility for the investigation of passive core cooling systems. In particular, a core make-up tank (CMT) has been installed in the rig to operate in a similar manner to those in many Advanced PWR designs. Three small break tests, GDE-24, GDE-34 and GDE-43 in the PACTEL facility were chosen for modelling with RELAP5. The objective of GDE-24 was to investigate CMT behaviour and in particular the effects of condensation in the CMT. The second test, GDE-34, was similar except that it had a smaller CMT and at the start of the test the water in the CMT and connecting pipework was at an elevated temperature. Test GDE-43 focused on conditions when the driving force for flow through the passive system injection system (PSIS) slowly disappears. Analysis of all tests reported here was carried out with RELAP5/MOD 3.2.1.2. The paper summarises the conclusions of all the tests. A critical part of the study revolved around modelling of the CMT. A model was developed to allow its detailed behaviour to be investigated more easily. This enabled recommendations for improving the condensation modelling in RELAP5 to be made. Apart from the wall condensation modelling issue, the implication of the work is that RELAP5/MOD 3.2.1.2 (a comparatively recent version of the code) is broadly adequate for these applications. (author)

Passive decay heat removal from the core region

International Nuclear Information System (INIS)

Hichen, E.F.; Jaegers, H.

2002-01-01

The decay heat in commercial Light Water Reactors is commonly removed by active and redundant safety systems supported by emergency power. For advanced power plant designs passive safety systems using a natural circulation mode are proposed: several designs are discussed. New experimental data gained with the NOKO and PANDA facilities as well as operational data from the Dodewaard Nuclear Power Plant are presented and compared with new calculations by different codes. In summary, the effectiveness of these passive decay heat removal systems have been demonstrated: original geometries and materials and for the NOKO facility and the Dodewaard Reactor typical thermal-hydraulic inlet and boundary conditions have been used. With several codes a good agreement between calculations and experimental data was achieved. (author)

Measurement of electrical impedance of a Berea sandstone core during the displacement of saturated brine by oil and CO2 injections

Science.gov (United States)

Liu, Yu; Xue, Ziqiu; Park, Hyuck; Kiyama, Tamotsu; Zhang, Yi; Nishizawa, Osamu; Chae, Kwang-seok

2015-12-01

Complex electrical impedance measurements were performed on a brine-saturated Berea sandstone core while oil and CO2 were injected at different pressures and temperatures. The saturations of brine, oil, and CO2 in the core were simultaneously estimated using an X-ray computed tomography scanner. The formation factor of this Berea core and the resistivity indexes versus the brine saturations were calculated using Archie's law. The experimental results found different flow patterns of oil under different pressures and temperatures. Fingers were observed for the first experiment at 10 MPa and 40 °C. The fingers were restrained as the viscosity ratio of oil and water changed in the second (10 MPa and 25 °C) and third (5 MPa and 25 °C) experiments. The resistivity index showed an exponential increase with a decrease in brine saturation. The saturation exponent varied from 1.4 to 4.0 at different pressure and temperature conditions. During the oil injection procedure, the electrical impedance increased with oil saturation and was significantly affected by different oil distributions; therefore, the impedance varied whether the finger was remarkable or not, even if the oil saturation remained constant. During the CO2 injection steps, the impedance showed almost no change with CO2 saturation because the brine in the pores became immobile after the oil injection.

The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

International Nuclear Information System (INIS)

Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

2015-01-01

Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

A passive solar heater-refrigerator

International Nuclear Information System (INIS)

D'Isep, F.; Sertorio, L.

1983-01-01

In this paper it is studied the nonequilibrium thermodynamic steady-state behaviour of a model system representing a core surrounded by an envelope in which the envelope interacts with the solar radiation and with an external bath having a given temperature profile. The heat flow between core and envelope can be controlled by varying the thermal conductivity of their interface. It is shown that this system acts as a passive heat pump raising the core average temperature with respect to the average equilibrium value corresponding to a fixed value of the interface conductivity, at the same time flattening its oscillation in time. By changing the time dependence of the conductivity the system vice versa acts as a refrigerator. It is shown how the limits of this performance depend on the passive parameters such as surfaces, conductivities, heat capacities. The periodicity considered in this study is the daily cycle

Optimization of the injection molding process for development of high performance calcium oxide -based ceramic cores

Science.gov (United States)

Zhou, P. P.; Wu, G. Q.; Tao, Y.; Cheng, X.; Zhao, J. Q.; Nan, H.

2018-02-01

The binder composition used for ceramic injection molding plays a crucial role on the final properties of sintered ceramic and to avoid defects on green parts. In this study, the effects of binder compositions on the rheological, microstructures and the mechanical properties of CaO based ceramic cores were investigated. It was found that the optimized formulation for dispersant, solid loading was 1.5 wt% and 84 wt%, respectively. The microstructures, such as porosity, pore size distribution and grain boundary density were closely related to the plasticizer contents. The decrease of plasticizer contents can enhance the strength of the ceramic cores but with decreased shrinkage. Meanwhile, the creep resistance of ceramic cores was enhanced by decreasing of plasticizer contents. The flexural strength of the core was found to decrease with the increase of the porosity, the improvement of creep resistance is closely related to the decrease of porosity and grain boundary density.

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

2015-09-15

Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

Passivation Layers for Indoor Solar Cells at Low Irradiation Intensities

OpenAIRE

Rühle, K.; Rauer, M.; Rüdiger, M.; Giesecke, J.; Niewelt, T.; Schmiga, C.; Glunz, S.W.; Kasemann, M.

2012-01-01

The passivation mechanisms and qualities of Al2O3, SiNx, SiO2 and a-Si:H(i) on p- and n-type silicon are investigated by quasi-steady-state photoluminescence measurements. This technique allows effective lifetime measurements in an extremely large injection range between 1010 cm-3 and 1017 cm-3. The measurements are discussed focusing on injections below 1012 cm-3 in order to determine the most effective passivation layer for solar cells arranged for indoor applications. Fixed negative charge...

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.

Passive safety systems for decay heat removal of MRX

Energy Technology Data Exchange (ETDEWEB)

Ochiai, M; Iida, H; Hoshi, T [Japan Atomic Energy Research Inst., Ibaraki (Japan). Nuclear Ship System Lab.

1996-12-01

The MRX (marine Reactor X) is an advanced marine reactor, its design has been studied in Japan Atomic Energy Research Institute. It is characterized by four features, integral type PWR, in-vessel type control rod drive mechanisms, water-filled containment vessel and passive decay heat removal system. A water-filled containment vessel is of great advantage since it ensures compactness of a reactor plant by realizing compact radiation shielding. The containment vessel also yields passive safety of MRX in the event of a LOCA by passively maintaining core flooding without any emergency water injection. Natural circulation of water in the vessels (reactor and containment vessels) is one of key factors of passive decay heat removal systems of MRX, since decay heat is transferred from fuel rods to atmosphere by natural circulation of the primary water, water in the containment vessel and thermal medium in heat pipe system for the containment vessel water cooling in case of long terms cooling after a LOCA as well as after reactor scram. Thus, the ideal of water-filled containment vessel is considered to be very profitable and significant in safety and economical point of view. This idea is, however, not so familiar for a conventional nuclear system, so experimental and analytical efforts are carried out for evaluation of hydrothermal behaviours in the reactor pressure vessel and in the containment vessel in the event of a LOCA. The results show the effectiveness of the new design concept. Additional work will also be conducted to investigate the practical maintenance of instruments in the containment vessel. (author). 4 refs, 9 figs, 2 tabs.

Design process of the nanofluid injection mechanism in nuclear power plants

Directory of Open Access Journals (Sweden)

Bang In Choel

2011-01-01

Full Text Available Abstract Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs. This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

French concepts of ''passive safety''

International Nuclear Information System (INIS)

Dennielou, Y.; Serret, M.

1990-01-01

N 4 model, the French 1400 MW PWR of the 90's, exhibits many advanced features. As far as safety is concerned, the fully computerized control room design takes advantage of the operating experience feedback and largely improves the man machine interface. New post-accident procedures have been developed (the so-called ''physical states oriented procedures''). A complete consistent set of ''Fundamental Safety Rules'' have been issued. This however doesn't imply any significant modification of standard PWR with regard to the passive aspects of safety systems or functions. Nevertheless, traditional PWR safety systems largely use passive aspects: natural circulation, reactivity coefficients, gravity driven control rods, injection accumulators, so on. Moreover, probability calculations allow for comparison between the respective contributions of passive and of active failures. In the near future, eventual options of future French PWRs to be commissioned after 2000 will be evaluated; simplification, passive and forgiving aspects of safety systems will be thoroughly considered. (author)

Hualong One's nuclear reactor core design and relative safety issues research

Energy Technology Data Exchange (ETDEWEB)

Yu, H., E-mail: yuhong_xing@126.com [Nuclear Power Inst. of China, Design and Research Sub-Inst., Chengdu, Sichuan (China)

2015-07-01

'Full text:' Hualong One, a third generation 1000MWe-class pressurized water reactor, is developed by China National Nuclear Cooperation (CNNC), based on the self-reliant technologies and experiences from China 40 years designing, construction, operation and maintenance of NPPs. In China, it has been approved to construct at Fuqing 5&6 and Fangchenggang 3&4. The Hualong One adopts advanced design features to dramatically enhance plant safety, economic efficiency and convenience of operation and maintenance. It consists of three loops with nominal thermal power output 3060 MWt and a 60-year design life. Its reactor core has 177 fuel assemblies, 18 month refueling interval (after initial cycle), and more than 15% thermal margin. It adopts low leakage loading pattern which can achieve better economy of the neutron, higher reactivity and lower radiation damage of pressure vessel. For the safety design, incorporating the feedback of Fukushima accident, the Hualong One has a combination of active and passive safety systems, a single station layout, double containment structure, and comprehensive implementation of defence-in-depth design principles. The new design features has been successfully evaluated to ensure that they enhance the performance and safety of Hualong One. Several experimental activates have been conducted, such as cavity injection and cooling system testing, passive containment heat removal system testing, and passive residual heat removal system of secondary side testing. The future improvements of Hualong reactor will focus on better economic core design and more reliable safety system. (author)

Assessment of MARS for Direct Contact Condensation in the Core Make-up Tank

International Nuclear Information System (INIS)

Park, Keun Tae; Park, Ik Kyu; Lee, Seung Wook

2013-01-01

In order to improve safety features under loss of coolant accident (LOCA) conditions, in many advanced light water reactors, gravity driven passive safety injection systems (PSISs) replace active pump driven emergency core cooling systems. Among various PSISs, the core make-up tank (CMT) with the pressure balancing line (PBL) and the coolant injection line (IL) represents an effective means of providing core cooling. Because the fluid is always sensing the reactor coolant system (RCS) through the PBL connecting the inlet of the CMT to the pressurizer in the case of CP1300 or to the cold legs in the case of AP600/1000, the CMT can provide cold water at any RCS pressure by gravity force. However, after the initiation of LOCAs, if the injection (or isolation) valve is opened, and the steam from the RCS is jetting into the highly subcooled liquid in the CMT and the enhanced interfacial area results in rapid condensation, which in turn, causes a rapid pressure drop in the CMT. As a result, the CMT pressure becomes less than the RCS pressure, and the injection of the CMT can be delayed until the CMT pressure builds up due to greatly reduced condensation in the CMT by the thermal stratification. In order to identify the parameters having significant effects on the gravity-driven injection and the major condensation modes, Lee and No (1998) conducted the separated effect tests of CMT with a small-scale facility. MARS has been developed as a multi-dimensional thermal-hydraulic (TH) system analysis code for the realistic simulation of two-phase TH transients for pressurized water reactor plants. As the backbones for the MARS code, the RELAP5/MOD3.2 and the COB-RA-TF codes were adopted. Recently, Chun et al. (2013) evaluated performance of the SMART passive safety system for SBLOCA using MARS code. However, it is not clarified that MARS can simulate properly the direct contact condensation in the CMT. Thus, in this study, we assess the analysis capability of the MARS code for

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Facile Phosphine-Free Synthesis of CdSe/ZnS Core/Shell Nanocrystals Without Precursor Injection

Directory of Open Access Journals (Sweden)

Zhu Chang-Qing

2008-01-01

Full Text Available AbstractA new simple method for synthesis of core/shell CdSe/ZnS nanocrystals (NCs is present. By adapting the use of cadmium stearate, oleylamine, and paraffin liquid to a non-injection synthesis and by applying a subsequent ZnS shelling procedure to CdSe NCs cores using Zinc acetate dihydrate and sulfur powder, luminescent CdSe/ZnS NCs with quantum yields of up to 36% (FWHM 42–43 nm were obtained. A seeding-growth technique was first applied to the controlled synthesis of ZnS shell. This method has several attractive features, such as the usage of low-cost, green, and environmentally friendlier reagents and elimination of the need for air-sensitive, toxic, and expensive phosphines solvent. Furthermore, due to one-pot synthetic route for CdSe/ZnS NCs, the approach presented herein is accessible to a mass production of these NCs.

Considerations on nuclear reactor passive safety systems

International Nuclear Information System (INIS)

2016-01-01

After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

Two-wavelength, passive self-injection-controlled operation of diode-pumped cw Yb-doped crystal lasers.

Science.gov (United States)

Louyer, Yann; Wallerand, Jean-Pierre; Himbert, Marc; Deneva, Margarita; Nenchev, Marin

2003-09-20

We demonstrate and investigate a peculiar mode of cw Yb3+-doped crystal laser operation when two emissions, at two independently tunable wavelengths, are simultaneously produced. Both emissions are generated from a single pumped volume and take place in either a single beam or spatially separated beams. The laser employs original two-channel cavities that use a passive self-injection-locking (PSIL) control to reduce intracavity loss. The advantages of the application of the PSIL technique and some limitations are shown. The conditions for two-wavelength multimode operation of the cw quasi-three-level diode-pumped Yb3+ lasers and the peculiarity of such an operation are carried out both theoretically and experimentally. The results reported are based on the example of a Yb3+:GGG laser but similar results are also obtained with a Yb3+:YAG laser. The laser operates in the 1023-1033-nm (1030-1040-nm) range with a total output power of 0.4 W. A two-wavelength, single longitudinal mode generation is also obtained.

Water injection device for reactor container

International Nuclear Information System (INIS)

Sakaki, Isao.

1996-01-01

A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

Preliminary study on functional performance of compound type multistage safety injection tank

International Nuclear Information System (INIS)

Bae, Youngmin; Kim, Young In; Kim, Keung Koo

2015-01-01

Highlights: • Functional performance of compound type multistage safety injection tanks is studied. • Effects of key design parameters are scrutinized. • Distinctive flow features in compound type safety injection tanks are explored. - Abstract: A parametric study is carried out to evaluate the functional performance of a compound type multistage safety injection tank that would be considered one of the components for the passive safety injection systems in nuclear power plants. The effects of key design parameters such as the initial volume fraction and charging pressure of gas, tank elevation, vertical location of a sparger, resistance coefficient, and operating condition on the injection flow rate are scrutinized along with a discussion of the relevant flow features. The obtained results indicate that the compound type multistage safety injection tank can effectively control the injection flow rate in a passive manner, by switching the driving force for the safety injection from gas pressure to gravity during the refill and reflood phases, respectively

Is ultrasound-guided injection more effective in chronic subacromial bursitis?

Science.gov (United States)

Hsieh, Lin-Fen; Hsu, Wei-Chun; Lin, Yi-Jia; Wu, Shih-Hui; Chang, Kae-Chwen; Chang, Hsiao-Lan

2013-12-01

Although ultrasound (US)-guided subacromial injection has shown increased accuracy in needle placement, whether US-guided injection produces better clinical outcome is still controversial. Therefore, this study aimed to compare the efficacy of subacromial corticosteroid injection under US guidance with palpation-guided subacromial injection in patients with chronic subacromial bursitis. Patients with chronic subacromial bursitis were randomized to a US-guided injection group and a palpation-guided injection group. The subjects in each group were injected with a mixture of 0.5 mL dexamethasone suspension and 3 mL lidocaine into the subacromial bursa. The primary outcome measures were the visual analog scale for pain and active and passive ranges of motion of the affected shoulder. Secondary outcome measures were the Shoulder Pain and Disability Index, the Shoulder Disability Questionnaire, and the 36-item Short-Form Health Survey (SF-36). The primary outcome measures were evaluated before, immediately, 1 wk, and 1 month after the injection; the secondary outcome measures were evaluated before, 1 wk, and 1 month after the injection. Of the 145 subjects screened, 46 in each group completed the study. Significantly greater improvement in passive shoulder abduction and in physical functioning and vitality scores on the SF-36 were observed in the US-guided group. The pre- and postinjection within-group comparison revealed significant improvement in the visual analog scale for pain and range of motion, as well as in the Shoulder Pain and Disability Index, Shoulder Disability Questionnaire, and SF-36 scores, in both groups. The US-guided subacromial injection technique produced significantly greater improvements in passive shoulder abduction and in some items of the SF-36. US is effective in guiding the needle into the subacromial bursa in patients with chronic subacromial bursitis.

Feasibility study on emergency passive habitability systems of SPWR

International Nuclear Information System (INIS)

Obata, H.; Tabata, H.; Urakami, M.; Naito, T.

2000-01-01

The major characteristic of the Simplified Pressurized Water Reactor (SPWR) is that safety systems for the emergency core cooling and the core decay heat removal functions are achieved by passive equipment. The AP600 developed in the U.S adopts passive emergency habitability system for the main control room (MCR) and the electrical equipment rooms (EER) by using the concrete of the structures as a heat sink. For the SPWR, alternative natural circulation cooling systems have been investigated: for MCR cooling, a cold water reservoir is used as heat sink; for EER cooling, outside air is instead employed. The distribution of the air-velocity and temperature in those rooms were calculated by using a three-dimensional thermal fluid analysis code. The authors verified the conceptual feasibility of these systems as the emergency passive habitability systems in the SPWR. (author)

Botulinum Neurotoxin A Injections Influence Stretching of the Gastrocnemius Muscle-Tendon Unit in an Animal Model

Directory of Open Access Journals (Sweden)

Christopher J. Tuohy

2012-08-01

Full Text Available Botulinum Neurotoxin A (BoNT-A injections have been used for the treatment of muscle contractures and spasticity. This study assessed the influence of (BoNT-A injections on passive biomechanical properties of the muscle-tendon unit. Mouse gastrocnemius muscle (GC was injected with BoNT-A (n = 18 or normal saline (n = 18 and passive, non-destructive, in vivo load relaxation experimentation was performed to examine how the muscle-tendon unit behaves after chemical denervation with BoNT-A. Injection of BoNT-A impaired passive muscle recovery (15% vs. 35% recovery to pre-stretching baseline, p < 0.05 and decreased GC stiffness (0.531 ± 0.061 N/mm vs. 0.780 ± 0.037 N/mm, p < 0.05 compared to saline controls. The successful use of BoNT-A injections as an adjunct to physical therapy may be in part attributed to the disruption of the stretch reflex; thereby modulating in vivo passive muscle properties. However, it is also possible that BoNT-A injection may alter the structure of skeletal muscle; thus modulating the in vivo passive biomechanical properties of the muscle-tendon unit.

Transient performance analysis of pressurized safety injection tank with a partition

International Nuclear Information System (INIS)

Bae, Youngmin; Kim, Young In; Kim, Keung Koo

2015-01-01

Highlights: • Functional performance of safety injection tanks with a partition is evaluated. • Effects of key design parameters are scrutinized. • Distinctive features of the flow in multi-unit safety injection tanks are explored. - Abstract: A parametric study has been performed to evaluate the functional performance of a pressurized multi-unit safety injection tank, which would be considered as one of the candidates for a passive safety injection system in a nuclear power plant. The influences of key design parameters including the orifice size, initial gas fraction, and resistance coefficients and operating condition on the injection flow rate are scrutinized with a discussion of the relevant flow features such as the choked flow of gas through an orifice and two interconnected regions of differing gaseous pressure. The obtained results indicate that a multi-unit safety injection tank can passively control the injection flow rate and provide a stable safety injection over a relatively long period even in the case of drastic depressurization of a reactor coolant system

NPR and ANSI Containment Study Using Passive Cooling Techniques

International Nuclear Information System (INIS)

Shin, J. J.; Iotti, R. C.; Wright, R. F.

1993-01-01

Passive containment cooling study of NPR (New Production Reactor) and ANSI (Advanced Neutron Source) following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally. All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified Contempt code Contempt-Npr. In order to verify proper heat transfer coefficient, temperature, heat flux and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and halt foot (1.7m) diameter outer cylinder. Comparison of Contempt-Npr and WGOTHIC was done for reduced scale Npr. It is concluded that Npr and ANSI containments can be passively cooled with air alone without extended cooling surfaces or passive water spray

Gaseous poison injection device

International Nuclear Information System (INIS)

Kubota, Ryuji; Sugisaki, Toshihiko; Inada, Ikuo.

1983-01-01

Purpose: To rapidly control the chain reaction due to thermal neutrons in a reactor core by using gaseous poisons as back-up means for control rod drives. Constitution: Gaseous poisons having a large neutron absorption cross section are used as back-up means for control rod drives. Upon failure of control rod insertion, the gaseous poisons are injected into the lower portion of the reactor core to control the reactor power. As the gaseous poisons, vapors at a high temperature and a higher pressure than that of the coolants in the reactor core are injected to control the reactor power due to the void effects. Since the gaseous poisons thus employed rapidly reach the reactor core and form gas bubbles therein, the deccelerating effect of the thermal neutrons is decreased to reduce the chain reaction. (Moriyama, K.)

Approaches to passive safety in advanced thermal reactors

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

Two types of a passive safety containment for a near future BWR with active and passive safety systems

International Nuclear Information System (INIS)

Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

2009-01-01

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

Two types of a passive safety containment for a near future BWR with active and passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Sato, Takashi [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)], E-mail: takashi44.sato@glb.toshiba.co.jp; Akinaga, Makoto; Kojima, Yoshihiro [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)

2009-09-15

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

Environmental mitigation for SCC initiation of BWR core internals by hydrogen injection during start-up

International Nuclear Information System (INIS)

Dozaki, K.; Abe, A.; Nagata, N.; Takiguchi, H.

2004-01-01

Hydrogen injection into the reactor water has been applied to many BWR power stations. Since hydrogen injected accelerates recombination of oxidant generated by water radiolysis, oxidant concentration, such as dissolved oxygen concentration in reactor water can be reduced. As the result of the reduction of oxidant concentration, Electrochemical Corrosion Potential (ECP) at the surface of structural material can be lowered. Lowered ECP moderates Stress Corrosion Cracking (SCC) sensitivity of structural materials, such as stainless steels. As usual, hydrogen injection system begins to work after the plant start-up is finished, when the condition of normal operation is established. Accordingly, Hydrogen Water Chemistry (HWC) does not cover all the period of plant operation. As far as SCC crack growth is considered, loss of HWC during plant start-up does not result in significant crack growth, because of duration of plant start-up is much shorter than that of plant normal operation, when HWC condition is being satisfied. However, the reactor water environment and load conditions during a plant start-up may contribute to the initiation of SCC. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Dissolved oxygen (DO) and hydrogen peroxide (H 2 O 2 ) has a peak, and ECP is in high levels during start-up. Therefore it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation. We call it HWC During Start-up (HDS) here. (orig.)

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

A PIV Study of Slotted Air Injection for Jet Noise Reduction

Science.gov (United States)

Henderson, Brenda S.; Wernet, Mark P.

2012-01-01

Results from acoustic and Particle Image Velocimetry (PIV) measurements are presented for single and dual-stream jets with fluidic injection on the core stream. The fluidic injection nozzles delivered air to the jet through slots on the interior of the nozzle at the nozzle trailing edge. The investigations include subsonic and supersonic jet conditions. Reductions in broadband shock noise and low frequency mixing noise were obtained with the introduction of fluidic injection on single stream jets. Fluidic injection was found to eliminate shock cells, increase jet mixing, and reduce turbulent kinetic energy levels near the end of the potential core. For dual-stream subsonic jets, the introduction of fluidic injection reduced low frequency noise in the peak jet noise direction and enhanced jet mixing. For dual-stream jets with supersonic fan streams and subsonic core streams, the introduction of fluidic injection in the core stream impacted the jet shock cell structure but had little effect on mixing between the core and fan streams.

Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

2014-01-01

Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

Thermophysiological adaptations to passive mild heat acclimation

NARCIS (Netherlands)

Pallubinsky, H; Schellen, L; Kingma, B R M; Dautzenberg, B; van Baak, M A; van Marken Lichtenbelt, W D

Passive mild heat acclimation (PMHA) reflects realistic temperature challenges encountered in everyday life. Active heat acclimation, combining heat exposure and exercise, influences several important thermophysiological parameters; for example, it decreases core temperature and enhances heat

Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

Energy Technology Data Exchange (ETDEWEB)

Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

2017-03-15

Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

Concept Design of a Gravity Core Cooling Tank as a Passive Residual Heat Removal System for a Research Reactor

International Nuclear Information System (INIS)

Lee, Kwonyeong; Chi, Daeyoung; Kim, Seong Hoon; Seo, Kyoungwoo; Yoon, Juhyeon

2014-01-01

A core downward flow is considered to use a plate type fuel because it is benefit to install the fuel in the core. If a flow inversion from a downward to upward flow in the core by a natural circulation is introduced within a high heat flux region of residual heat, the fuel fails instantly due to zero flow. Therefore, the core downward flow should be sufficiently maintained until the residual heat is in a low heat flux region. In a small power research reactor, inertia generated by a flywheel of the PCP can maintain a downward flow shortly and resolve the problem of a flow inversion. However, a high power research reactor more than 10 MW should have an additional method to have a longer downward flow until a low heat flux. Usually, other research reactors have selected an active residual heat removal system as a safety class. But, an active safety system is difficult to design and expensive to construct. A Gravity Core Cooling Tank (GCCT) beside the reactor pool with a Residual Heat Removal Pipe connecting two pools was developed and designed preliminarily as a passive residual heat removal system for an open-pool type research reactor. It is very simple to design and cheap to construct. Additionally, a non-safety, but active residual heat removal system is applied with the GCCT. It is a Pool Water Cooling and Purification System. It can improve the usability of the research reactor by removing the thermal waves, and purify the reactor pool, the Primary Cooling System, and the GCCT. Moreover, it can reduce the pool top radiation level

Engineered safeguards and passive safety features (safety analysis detailed report no. 6)

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The Safety-Analysis Summary lists the reactor's safety aspects for passive and active prevention of severe accidents and mitigation of accident consequences, i.e., intrinsic and passive protections of the plant; intrinsic and passive protections of the core; inherent decay-heat removal systems; rapid-shutdown systems; four physical containment barriers. This report goes into further details regarding some of this aspects.

Analysis of large break loss of coolant accident with simultaneous injection into cold leg and hot leg

International Nuclear Information System (INIS)

Luo Bangqi

1997-01-01

When a large break loss of coolant accident occurs, the most part of the safety injection water injected into the cold leg by the safety injection system will flow through the channel between the pressure vessel and the barrel out of the break into the containment, only a little part of the safety injection water can flow into the reactor core. If the safety injection can inject into both the cold leg and the hot leg simultaneously, the safety injection water injected from the cold leg will flow into the core more easily, because the safety injection water injected from the hot leg will carry out more heat from the upper plenum and the core, so the upper plenum and the core is depressed. In addition, a small part of the safety injection water injected from the hot leg will flow down in the core after impinging the guide tubes in the upper plenum, so the core will get more safety injection water than only cold leg injection, and the core will be much safer

Hydrogen injection device in BWR type reactor

International Nuclear Information System (INIS)

Takagi, Jun-ichi; Kubo, Koji.

1988-01-01

Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

Many advanced light water reactor designs incorporate passive rather than active safety features for front-line accident response. A method for evaluating the reliability of these passive systems in the context of probabilistic risk assessment has been developed at Sandia National Laboratories. This method addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. These processes provide the system's driving force; examples are natural circulation and gravity-induced injection. This paper describes the method, and provides some preliminary results of application of the approach to the Westinghouse AP600 design

Safety characteristics of the US advanced liquid metal reactor core

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

1991-01-01

The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

Performance Verification for Safety Injection Tank with Fluidic Device

International Nuclear Information System (INIS)

Yune, Seok Jeong; Kim, Da Yong

2014-01-01

In LBLOCA, the SITs of a conventional nuclear power plant deliver excessive cooling water to the reactor vessel causing the water to flow into the containment atmosphere. In an effort to make it more efficient, Fluidic Device (FD) is installed inside a SIT of Advanced Power Reactor 1400 (APR 1400). FD, a complete passive controller which doesn't require actuating power, controls injection flow rates which are susceptible to a change in the flow resistance inside a vortex chamber of FD. When SIT Emergency Core Cooling (ECC) water level is above the top of the stand pipe, the water enters the vortex chamber through both the top of the stand pipe and the control ports resulting in injection of the water at a large flow rate. When the water level drops below the top of the stand pipe, the water only enters the vortex chamber through the control ports resulting in vortex formation in the vortex chamber and a relatively small flow injection. Performance verification of SIT shall be carried out because SITs play an integral role to mitigate accidents. In this paper, the performance verification method of SIT with FD is presented. In this paper, the equations for calculation of flow resistance coefficient (K) are induced to evaluate on-site performance of APR 1400 SIT with FD. Then, the equations are applied to the performance verification of SIT with FD and good results are obtained

Westinghouse Small Modular Reactor passive safety system response to postulated events

International Nuclear Information System (INIS)

Smith, M. C.; Wright, R. F.

2012-01-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 R reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

International Nuclear Information System (INIS)

Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2015-01-01

Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

Acute lethal toxicity following passive immunization for treatment of murine cryptococcosis.

OpenAIRE

Savoy, A C; Lupan, D M; Manalo, P B; Roberts, J S; Schlageter, A M; Weinhold, L C; Kozel, T R

1997-01-01

Passive immunization with monoclonal antibodies (MAbs) specific for the major capsular polysaccharide of Cryptococcus neoformans alters the course of murine cryptococcosis. During studies of passive immunization for treatment of murine cryptococcosis, we noted the occurrence of an acute, lethal toxicity. Toxicity was characterized by scratching, lethargy, respiratory distress, collapse, and death within 20 to 60 min after injection of antibody. The toxic effect was observed only in mice with ...

AC-600 passive ECRHR system and its research program

International Nuclear Information System (INIS)

Chen Bingde; Xiao Zejun; Zhou Renmin; Liu Yiyang

1997-01-01

The secondary-side passive emergency core residual heat removal system (ECRHR System) is an important part of AC-600 PWR passive safety system, with which the core decay heat can be removed through nature circulation in primary and secondary system. Since 1991, the program for AC-600 passive ECRHR system has been conducted to investigate its distinct thermal-hydraulic phenomena, heat removal capability, affecting factors, and to develop computer codes. The test facility, designed according to the power/volume simulating law, is a full pressure and temperature operating loop with volume scaling factor of 1/390. It is composed of main loop system, emergence feedwater system, depression system, heat tracing, I and C system and power supply system. A total of sixteen tests is planned in first stage and fifteen of them have been done. The preliminary result analysis showed that the system has efficient heat removal capability in most conditions and some special thermal hydraulic phenomena, for example, flow fluctuation, which has negative impact on system's nature circulation, were identified

The passive system for reflooding of the VVER reactor core from the second-stage hydro-accumulators: design and basic design solutions

International Nuclear Information System (INIS)

Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko; Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev

2005-01-01

Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m 3 . The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)

Active and passive spectroscopic imaging in the DIII-D tokamak

International Nuclear Information System (INIS)

Van Zeeland, M A; Brooks, N H; Burrell, K H; Groebner, R J; Hyatt, A W; Luce, T C; Wade, M R; Yu, J H; Pablant, N; Heidbrink, W W; Solomon, W M

2010-01-01

Wide-angle, 2D imaging of Doppler-shifted, Balmer alpha (D α ) emission from high energy injected neutrals, charge exchange recombination (CER) emission from neutral beam interaction with thermal ions and fully stripped impurity ions and visible bremsstrahlung (VB) from the core of DIII-D plasmas has been carried out. Narrowband interference filters were used to isolate the specific wavelength ranges of visible radiation for detection by a tangentially viewing, fast-framing camera. Measurements of the D α emission from fast neutrals injected into the plasma from the low field side reveal the vertical distribution of the beam, its divergence and the variation in its radial penetration with density. Modeling of this emission using both a full Monte Carlo collisional radiative code as well as a simple beam attenuation code coupled to Atomic Data and Analysis Structure emissivity lookup tables yields qualitative agreement, however the absolute magnitudes of the emissivities in the predicted distribution are larger than those measured. Active measurements of carbon CER brightness are in agreement with those made independently along the beam midplane using DIII-D's multichordal, CER spectrometer system, confirming the potential of this technique for obtaining 2D profiles of impurity density. Passive imaging of VB, which can be inverted to obtain local emissivity profiles, is compared with measurements from both a calibrated filter/photomultiplier array and the standard multichordal CER spectrometer system.

Triggering Collapse of the Presolar Dense Cloud Core and Injecting Short-lived Radioisotopes with a Shock Wave. V. Nonisothermal Collapse Regime

Energy Technology Data Exchange (ETDEWEB)

Boss, Alan P., E-mail: aboss@carnegiescience.edu [Department of Terrestrial Magnetism, Carnegie Institution for Science, 5241 Broad Branch Road, NW, Washington, DC 20015-1305 (United States)

2017-08-01

Recent meteoritical analyses support an initial abundance of the short-lived radioisotope (SLRI) {sup 60}Fe that may be high enough to require nucleosynthesis in a core-collapse supernova, followed by rapid incorporation into primitive meteoritical components, rather than a scenario where such isotopes were inherited from a well-mixed region of a giant molecular cloud polluted by a variety of supernovae remnants and massive star winds. This paper continues to explore the former scenario, by calculating three-dimensional, adaptive mesh refinement, hydrodynamical code (FLASH 2.5) models of the self-gravitational, dynamical collapse of a molecular cloud core that has been struck by a thin shock front with a speed of 40 km s{sup −1}, leading to the injection of shock front matter into the collapsing cloud through the formation of Rayleigh–Taylor fingers at the shock–cloud intersection. These models extend the previous work into the nonisothermal collapse regime using a polytropic approximation to represent compressional heating in the optically thick protostar. The models show that the injection efficiencies of shock front materials are enhanced compared to previous models, which were not carried into the nonisothermal regime, and so did not reach such high densities. The new models, combined with the recent estimates of initial {sup 60}Fe abundances, imply that the supernova triggering and injection scenario remains a plausible explanation for the origin of the SLRIs involved in the formation of our solar system.

Characterization of tetraethylene glycol passivated iron nanoparticles

International Nuclear Information System (INIS)

Nunes, Eloiza da Silva; Viali, Wesley Renato; Silva, Sebastião William da; Coaquira, José Antonio Huamaní; Garg, Vijayendra Kumar; Oliveira, Aderbal Carlos de; Morais, Paulo César; Jafelicci Júnior, Miguel

2014-01-01

Graphical abstract: - Highlights: • Metallic iron nanoparticles were passivated in tetraethylene glycol media. • Passivated nanoparticles presented pomegranate-like core@shell structure. • Passivation of metallic iron correlates with the tetraethylene glycol degradation. • Boron enriched metallic iron phase was more susceptible to oxidation. • The iron oxide shell was identified as Fe 3 O 4 with a mass fraction of 43:53 related to αFe. - Abstract: The present study describes the synthesis and characterization of iron@iron oxide nanoparticles produced by passivation of metallic iron in tetraethylene glycol media. Structural and chemical characterizations were performed using transmission electron microscopy (TEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), and Mössbauer spectroscopy. Pomegranate-like core@shell nanoparticulate material in the size range of 90–120 nm was obtained. According to quantitative phase analysis using Rietveld structure refinement the synthesized iron oxide was identified as magnetite (Fe 3 O 4 ) whereas the iron to magnetite mass fractions was found to be 47:53. These findings are in good agreement with the data obtained from Mössbauer and thermal gravimetric analysis (TGA). The XPS data revealed the presence of a surface organic layer with higher hydrocarbon content, possibly due to the tetraethylene glycol thermal degradation correlated with iron oxidation. The room-temperature (300 K) saturation magnetization measured for the as-synthesized iron and for the iron–iron oxide were 145 emu g −1 and 131 emu g −1 , respectively. The measured saturation magnetizations are in good agreement with data obtained from TEM, XRD and Mössbauer spectroscopy

Fe-based nanocrystalline powder cores with ultra-low core loss

Energy Technology Data Exchange (ETDEWEB)

Wang, Xiangyue, E-mail: wangxiangyue1986@163.com [China Iron and Steel Research Institute Group, Beijing 100081 (China); Center of Advanced Technology and Materials Co., Ltd., Beijing 100081 (China); Lu, Zhichao; Lu, Caowei; Li, Deren [China Iron and Steel Research Institute Group, Beijing 100081 (China); Center of Advanced Technology and Materials Co., Ltd., Beijing 100081 (China)

2013-12-15

Melt-spun amorphous Fe{sub 73.5}Cu{sub 1}Nb{sub 3}Si{sub 15.5}B{sub 7} alloy strip was crushed to make flake-shaped fine powders. The passivated powders by phosphoric acid were mixed with organic and inorganic binder, followed by cold compaction to form toroid-shaped bonded powder-metallurgical magnets. The powder cores were heat-treated to crystallize the amorphous structure and to control the nano-grain structure. Well-coated phosphate-oxide insulation layer on the powder surface decreased the the core loss with the insulation of each powder. FeCuNbSiB nanocrystalline alloy powder core prepared from the powder having phosphate-oxide layer exhibits a stable permeability up to high frequency range over 2 MHz. Especially, the core loss could be reduced remarkably. At the other hand, the softened inorganic binder in the annealing process could effectively improve the intensity of powder cores. - Highlights: • Fe-based nanocrystalline powder cores were prepared with low core loss. • Well-coated phosphate-oxide insulation layer on the powder surface decreased the core loss. • Fe-based nanocrystalline powder cores exhibited a stable permeability up to high frequency range over 2 MHz. • The softened inorganic binder in the annealing process could effectively improve the intensity of powder cores.

Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

International Nuclear Information System (INIS)

Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

1990-07-01

In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

Core concepts for ''zero-sodium-void-worth core'' in metal fuelled fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

1991-01-01

Core design options to reduce the sodium void worth in metal fueled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a ''pancaked'' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. 16 refs., 2 figs., 3 tabs

Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

1991-01-01

Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

Scaling for Mixed Convection Heat Transfer in Passive Containments and Experiment Design

International Nuclear Information System (INIS)

Wang, Shengfei; Yu, Yu; Lv, Xuefeng; Niu, Fenglei; Yan, Xiuping

2012-01-01

Most of the advanced nuclear reactor design utilizes passive systems to remove heat from the core by natural circulation. The passive systems will be widely used in generation III pressurized water reactor. One of the typical passive systems is passive containment cooling system (PCCS), which is a passive condenser system designed to remove heat from the containment for long term cooling after a postulated reactor accident. In order to establish empirical correlations and develop simulation models, a scaling analysis is performed in designing an experiment for the prototype PCCS. This paper presents a scaling method and the design of the experimental facility. The key dimensionless parameters governing the dominant processes are given at last

The cognitive-behavioral system of leadership: cognitive antecedents of active and passive leadership behaviors

Science.gov (United States)

Dóci, Edina; Stouten, Jeroen; Hofmans, Joeri

2015-01-01

In the present paper, we propose a cognitive-behavioral understanding of active and passive leadership. Building on core evaluations theory, we offer a model that explains the emergence of leaders’ active and passive behaviors, thereby predicting stable, inter-individual, as well as variable, intra-individual differences in both types of leadership behavior. We explain leaders’ stable behavioral tendencies by their fundamental beliefs about themselves, others, and the world (core evaluations), while their variable, momentary behaviors are explained by the leaders’ momentary appraisals of themselves, others, and the world (specific evaluations). By introducing interactions between the situation the leader enters, the leader’s beliefs, appraisals, and behavior, we propose a comprehensive system of cognitive mechanisms that underlie active and passive leadership behavior. PMID:26441721

Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

International Nuclear Information System (INIS)

Wootan, D.W.; Omberg, R.P.

2015-01-01

Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

SWR 1000: an advanced boiling water reactor with passive safety features

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

Effects of Supercritical Environment on Hydrocarbon-fuel Injection

Institute of Scientific and Technical Information of China (English)

Bongchul Shin; Dohun Kim; Min Son; Jaye Koo

2017-01-01

In this study,the effects of environment conditions on decane were investigated.Decane was injected in subcritical and supercritical ambient conditions.The visualization chamber was pressurized to 1.68 MPa by using nitrogen gas at a temperature of 653 K for subcritical ambient conditions.For supercritical ambient conditions,the visualization chamber was pressurized to 2.52 MPa by using helium at a temperature of 653 K.The decane injection in the pressurized chamber was visualized via a shadowgraph technique and gradient images were obtained by a post processing method.A large variation in density gradient was observed at jet interface in the case of subcritical injection in subcritical ambient conditions.Conversely,for supercritical injection in supercritical ambient conditions,a small density gradient was observed at the jet interface.In a manner similar to that observed in other cases,supercritical injection in subcritical ambient conditions differed from supercritical ambient conditions such as sphere shape liquid.Additionally,there were changes in the interface,and the supercritical injection core width was thicker than that in the subcritical injection.Furthermore,in cases with the same injection conditions,the change in the supercritical ambient normalized core width was smaller than the change in the subcritical ambient normalized core width owing to high specific heat at the supercritical injection and small phase change at the interface.Therefore,the interface was affected by the changing ambient condition.Given that the effect of changing the thermodynamic properties of propellants could be essential for a variable thrust rocket engine,the effects of the ambient conditions were investigated experimentally.

Effects of supercritical environment on hydrocarbon-fuel injection

Science.gov (United States)

Shin, Bongchul; Kim, Dohun; Son, Min; Koo, Jaye

2017-04-01

In this study, the effects of environment conditions on decane were investigated. Decane was injected in subcritical and supercritical ambient conditions. The visualization chamber was pressurized to 1.68 MPa by using nitrogen gas at a temperature of 653 K for subcritical ambient conditions. For supercritical ambient conditions, the visualization chamber was pressurized to 2.52 MPa by using helium at a temperature of 653 K. The decane injection in the pressurized chamber was visualized via a shadowgraph technique and gradient images were obtained by a post processing method. A large variation in density gradient was observed at jet interface in the case of subcritical injection in subcritical ambient conditions. Conversely, for supercritical injection in supercritical ambient conditions, a small density gradient was observed at the jet interface. In a manner similar to that observed in other cases, supercritical injection in subcritical ambient conditions differed from supercritical ambient conditions such as sphere shape liquid. Additionally, there were changes in the interface, and the supercritical injection core width was thicker than that in the subcritical injection. Furthermore, in cases with the same injection conditions, the change in the supercritical ambient normalized core width was smaller than the change in the subcritical ambient normalized core width owing to high specific heat at the supercritical injection and small phase change at the interface. Therefore, the interface was affected by the changing ambient condition. Given that the effect of changing the thermodynamic properties of propellants could be essential for a variable thrust rocket engine, the effects of the ambient conditions were investigated experimentally.

An investigation into the effective surface passivation of quantum dots by a photo-assisted chemical method

Directory of Open Access Journals (Sweden)

So-Yeong Joo

2018-01-01

Full Text Available In this study, we have developed an effective amino passivation process for quantum dots (QDs at room temperature and have investigated a passivation mechanism using a photo-assisted chemical method. As a result of the reverse reaction of the H2O molecules, the etching kinetics of the photo-assisted chemical method increased upon increasing the 3-amino-1-propanol (APOL/H2O ratio of the etching solution. Photon-excited electron-hole pairs lead to strong bonding between the organic and surface atoms of the QDs, and results in an increase of the quantum yield (QY%. This passivation method is also applicable to CdSe/ZnSe core/shell structures of QDs, due to the passivation of mid-gap defects states at the interface. The QY% of the as-synthesized CdSe QDs is dramatically enhanced by the amino passivation from 37% to 75% and the QY% of the CdSe/ZnSe core/shell QDs is also improved by ∼28%.

SULEU NTP Core with Passive Reactivity Control and Enhanced Submersion Safety

Energy Technology Data Exchange (ETDEWEB)

Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Eades, Michael J [The Ohio State University, Ohio (United States)

2016-05-15

In this summary, SULEU has been adapted to implement some of the latest developments of LEUNTP design efforts. These include the implementation of a rapid depletion burnable absorber to flatten the reactivity profile during operation and the addition of a lower axial reflector to help minimize the reactivity increase during the full submersion criticality accident. The purpose of this study is to show the state of current LEU-NTP designs in terms of resolving key issues such as minimizing control drum usage and resolving the full submersion criticality accident. Future work will include integrating the rapid depletion poison with other passive reactivity control devices (such as hydrogen density in the tie-tubes) and developing additional systems for mitigating the full submersion criticality accident. It is widely acknowledged that nuclear thermal propulsion (NTP) is an enabling technology for manned missions to Mars and other locations beyond low-Earth orbit. Without nuclear thermal propulsion, manned space travel will be severely limited by the propellant requirements of chemical propulsion and significantly longer travel times of electric propulsion. While the performance superiority of NTP is clear, its implementation has been to date unsuccessful due to the significant costs of development, implementation, and regulations associated with the heritage NTP designs. These new systems take heritage designs and experimental results and adapt them to use LEU fuel with minimum impact on the heritage system. This is done in order to ensure their continued relevance with existing NTP research efforts and enable their rapid implementation into existing NASA efforts for human Mars mission planning. Of the current baseline NTP designs being studied, this paper concerns itself with the improvement of the Superb Use of Low Enriched (SULEU) core.

Trends vs. reactor size of passive reactivity shutdown and control performance

International Nuclear Information System (INIS)

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

1987-01-01

For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the massive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. For LMR concepts, the passive decay heat removal goal of inherent safety has been approached in US designs by use of pool layouts, larger surface to volume ratio of the reactor vessel with natural draft air cooling of the vessel surface, elevations and redans which promote natural circulation through the core, and thermal mass of the pool contents sufficient to absorb that initial transient decay heat which exceeds the natural draft air cooling capacity. This paper describes current US ''inherently safe'' reactor design

Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

International Nuclear Information System (INIS)

Su Xia

2013-01-01

The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

Passive heat removal in CANDU

International Nuclear Information System (INIS)

Hart, R.S.

1997-01-01

CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

Tuning of Passivity-Preserving Controllers for Switched-Mode Power Converters

NARCIS (Netherlands)

Jeltsema, Dimitri; Scherpen, Jacquelien M.A.

2004-01-01

Nonlinear passivity-based control (PBC) algorithms for power converters have proved to be an interesting alternative to other, mostly linear, control techniques. The control objective is usually achieved through an energy reshaping process and by injecting damping to modify the dissipation structure

Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

2016-05-15

CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

Characterization of tetraethylene glycol passivated iron nanoparticles

Energy Technology Data Exchange (ETDEWEB)

Nunes, Eloiza da Silva; Viali, Wesley Renato [Laboratório de Materiais Magnéticos e Coloides, Departamento de Físico-química, Instituto de Química, Universidade Estadual Paulista, Araraquara, SP 14801-970 (Brazil); Silva, Sebastião William da; Coaquira, José Antonio Huamaní; Garg, Vijayendra Kumar; Oliveira, Aderbal Carlos de [Instituto de Física, Núcleo de Física Aplicada, Universidade de Brasília, Brasília, DF 70910-900 (Brazil); Morais, Paulo César [Instituto de Física, Núcleo de Física Aplicada, Universidade de Brasília, Brasília, DF 70910-900 (Brazil); School of Automation, Huazhong University of Science and Technology, Wuhan 430074 (China); Jafelicci Júnior, Miguel, E-mail: jafeli@iq.unesp.br [Laboratório de Materiais Magnéticos e Coloides, Departamento de Físico-química, Instituto de Química, Universidade Estadual Paulista, Araraquara, SP 14801-970 (Brazil)

2014-10-01

Graphical abstract: - Highlights: • Metallic iron nanoparticles were passivated in tetraethylene glycol media. • Passivated nanoparticles presented pomegranate-like core@shell structure. • Passivation of metallic iron correlates with the tetraethylene glycol degradation. • Boron enriched metallic iron phase was more susceptible to oxidation. • The iron oxide shell was identified as Fe{sub 3}O{sub 4} with a mass fraction of 43:53 related to αFe. - Abstract: The present study describes the synthesis and characterization of iron@iron oxide nanoparticles produced by passivation of metallic iron in tetraethylene glycol media. Structural and chemical characterizations were performed using transmission electron microscopy (TEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), and Mössbauer spectroscopy. Pomegranate-like core@shell nanoparticulate material in the size range of 90–120 nm was obtained. According to quantitative phase analysis using Rietveld structure refinement the synthesized iron oxide was identified as magnetite (Fe{sub 3}O{sub 4}) whereas the iron to magnetite mass fractions was found to be 47:53. These findings are in good agreement with the data obtained from Mössbauer and thermal gravimetric analysis (TGA). The XPS data revealed the presence of a surface organic layer with higher hydrocarbon content, possibly due to the tetraethylene glycol thermal degradation correlated with iron oxidation. The room-temperature (300 K) saturation magnetization measured for the as-synthesized iron and for the iron–iron oxide were 145 emu g{sup −1} and 131 emu g{sup −1}, respectively. The measured saturation magnetizations are in good agreement with data obtained from TEM, XRD and Mössbauer spectroscopy.

Passive safety systems reliability and integration of these systems in nuclear power plant PSA

International Nuclear Information System (INIS)

La Lumia, V.; Mercier, S.; Marques, M.; Pignatel, J.F.

2004-01-01

Innovative nuclear reactor concepts could lead to use passive safety features in combination with active safety systems. A passive system does not need active component, external energy, signal or human interaction to operate. These are attractive advantages for safety nuclear plant improvements and economic competitiveness. But specific reliability problems, linked to physical phenomena, can conduct to stop the physical process. In this context, the European Commission (EC) starts the RMPS (Reliability Methods for Passive Safety functions) program. In this RMPS program, a quantitative reliability evaluation of the RP2 system (Residual Passive heat Removal system on the Primary circuit) has been realised, and the results introduced in a simplified PSA (Probabilistic Safety Assessment). The scope is to get out experience of definition of characteristic parameters for reliability evaluation and PSA including passive systems. The simplified PSA, using event tree method, is carried out for the total loss of power supplies initiating event leading to a severe core damage. Are taken into account: failures of components but also failures of the physical process involved (e.g. natural convection) by a specific method. The physical process failure probabilities are assessed through uncertainty analyses based on supposed probability density functions for the characteristic parameters of the RP2 system. The probabilities are calculated by MONTE CARLO simulation coupled to the CATHARE thermalhydraulic code. The yearly frequency of the severe core damage is evaluated for each accident sequence. This analysis has identified the influence of the passive system RP2 and propose a re-dimensioning of the RP2 system in order to satisfy the safety probabilistic objectives for reactor core severe damage. (authors)

Active stabilization of a diode laser injection lock.

Science.gov (United States)

Saxberg, Brendan; Plotkin-Swing, Benjamin; Gupta, Subhadeep

2016-06-01

We report on a device to electronically stabilize the optical injection lock of a semiconductor diode laser. Our technique uses as discriminator the peak height of the laser's transmission signal on a scanning Fabry-Perot cavity and feeds back to the diode current, thereby maintaining maximum optical power in the injected mode. A two-component feedback algorithm provides constant optimization of the injection lock, keeping it robust to slow thermal drifts and allowing fast recovery from sudden failures such as temporary occlusion of the injection beam. We demonstrate the successful performance of our stabilization method in a diode laser setup at 399 nm used for laser cooling of Yb atoms. The device eases the requirements on passive stabilization and can benefit any diode laser injection lock application, particularly those where several such locks are employed.

CFD Analysis of the Safety Injection Tank and Fluidic Device

Energy Technology Data Exchange (ETDEWEB)

Cho, Jai Oan; Nietiadi, Yohanes Setiawan; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Addad, Yacine [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-05-15

One of the most important components in the ECCS is the safety injection tank (SIT). Inside the SIT, a fluidic device is installed, which passively controls the mass flow of the safety injection and eliminates the need for low pressure safety injection pumps. As more passive safety mechanisms are being pursued, it has become more important to understand flow structure and the loss mechanism within the fluidic device. Current computational fluid dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study proposes to find a more exact result using CFD and more realistic modeling to predict the performance during accident scenarios more accurately. The safety injection tank with fluidic device was analyzed thoroughly using CFD. The preliminary calculation used 60,000 meshes for the initial test calculation. The results fit the experimental results surprisingly despite its coarse grid. Nonetheless, the mesh resolution was increased to capture the vortex in the fluidic device precisely. Once a detailed CFD computation is finished, a small-scale experiment will be conducted for the given conditions. Using the experimental results and the CFD model, physical models can be improved to fit the results more accurately.

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

International Nuclear Information System (INIS)

Scheuerer, Martina; Weis, Johannes

2012-01-01

Highlights: ► Pressurized thermal shocks are important phenomena for plant life extension and aging. ► The thermal-hydraulics of PTS have been studied experimentally and numerically. ► In the Large Scale Test Facility a loss of coolant accident was investigated. ► CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

Uncertainty reduction requirements in cores designed for passive reactivity shutdown

International Nuclear Information System (INIS)

Wade, D.C.

1988-01-01

The first purpose of this paper is to describe the changed focus of neutronics accuracy requirements existing in the current US advanced LMR development program where passive shutdown is a major design goal. The second purpose is to provide the background and rationale which supports the selection of a formal data fitting methodology as the means for the application of critical experiment measurements to meet these accuracy needs. 6 refs., 1 fig., 2 tabs

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Active and passive spectroscopic imaging in the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Van Zeeland, M A; Brooks, N H; Burrell, K H; Groebner, R J; Hyatt, A W; Luce, T C; Wade, M R [General Atomics, PO Box 85608 San Diego, CA 92186-5608 (United States); Yu, J H; Pablant, N [University of California-San Diego, 9500 Gilman Drive, La Jolla, CA 92093 (United States); Heidbrink, W W [University of California-Irvine, 4129 Frederick Reines Hall, Irvine, CA 92697 (United States); Solomon, W M, E-mail: vanzeeland@fusion.gat.co [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

2010-04-15

Wide-angle, 2D imaging of Doppler-shifted, Balmer alpha (D{sub a}lpha) emission from high energy injected neutrals, charge exchange recombination (CER) emission from neutral beam interaction with thermal ions and fully stripped impurity ions and visible bremsstrahlung (VB) from the core of DIII-D plasmas has been carried out. Narrowband interference filters were used to isolate the specific wavelength ranges of visible radiation for detection by a tangentially viewing, fast-framing camera. Measurements of the D{sub a}lpha emission from fast neutrals injected into the plasma from the low field side reveal the vertical distribution of the beam, its divergence and the variation in its radial penetration with density. Modeling of this emission using both a full Monte Carlo collisional radiative code as well as a simple beam attenuation code coupled to Atomic Data and Analysis Structure emissivity lookup tables yields qualitative agreement, however the absolute magnitudes of the emissivities in the predicted distribution are larger than those measured. Active measurements of carbon CER brightness are in agreement with those made independently along the beam midplane using DIII-D's multichordal, CER spectrometer system, confirming the potential of this technique for obtaining 2D profiles of impurity density. Passive imaging of VB, which can be inverted to obtain local emissivity profiles, is compared with measurements from both a calibrated filter/photomultiplier array and the standard multichordal CER spectrometer system.

Experimental and design experience with passive safety features of liquid metal reactors

International Nuclear Information System (INIS)

Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

1992-10-01

Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

Effect of long-term isometric training on core/torso stiffness.

Science.gov (United States)

Lee, Benjamin C Y; McGill, Stuart M

2015-06-01

Although core stiffness enhances athletic performance traits, controversy exists regarding the effectiveness of isometric vs. dynamic core training methods. This study aimed to determine whether long-term changes in stiffness can be trained, and if so, what is the most effective method. Twenty-four healthy male subjects (23 ± 3 years; 1.8 ± 0.06 m; 77.5 ± 10.8 kg) were recruited for passive and active stiffness measurements before and after a 6-week core training intervention. Twelve subjects (22 ± 2 years; 1.8 ± 0.08 m; 78.3 ± 12.3 kg) were considered naive to physical and core exercise. The other 12 subjects (24 ± 3 years; 1.8 ± 0.05 m; 76.8 ± 9.7 kg) were Muay Thai athletes (savvy). A repeated-measures design compared core training methods (isometric vs. dynamic, with a control group) and subject training experience (naive vs. savvy) before and after a 6-week training period. Passive stiffness was assessed on a "frictionless" bending apparatus and active stiffness assessed through a quick release mechanism. Passive stiffness increased after the isometric training protocol. Dynamic training produced a smaller effect, and as expected, there was no change in the control group. Active stiffness did not change in any group. Comparisons between subject and training groups did not reveal any interactions. Thus, an isometric training approach was superior in terms of enhancing core stiffness. This is important since increased core stiffness enhances load bearing ability, arrests painful vertebral micromovements, and enhances ballistic distal limb movement. This may explain the efficacy reported for back and knee injury reduction.

James Webb Space Telescope Core 2 Test - Cryogenic Thermal Balance Test of the Observatorys Core Area Thermal Control Hardware

Science.gov (United States)

Cleveland, Paul; Parrish, Keith; Thomson, Shaun; Marsh, James; Comber, Brian

2016-01-01

The James Webb Space Telescope (JWST), successor to the Hubble Space Telescope, will be the largest astronomical telescope ever sent into space. To observe the very first light of the early universe, JWST requires a large deployed 6.5-meter primary mirror cryogenically cooled to less than 50 Kelvin. Three scientific instruments are further cooled via a large radiator system to less than 40 Kelvin. A fourth scientific instrument is cooled to less than 7 Kelvin using a combination pulse-tube Joule-Thomson mechanical cooler. Passive cryogenic cooling enables the large scale of the telescope which must be highly folded for launch on an Ariane 5 launch vehicle and deployed once on orbit during its journey to the second Earth-Sun Lagrange point. Passive cooling of the observatory is enabled by the deployment of a large tennis court sized five layer Sunshield combined with the use of a network of high efficiency radiators. A high purity aluminum heat strap system connects the three instrument's detector systems to the radiator systems to dissipate less than a single watt of parasitic and instrument dissipated heat. JWST's large scale features, while enabling passive cooling, also prevent the typical flight configuration fully-deployed thermal balance test that is the keystone of most space missions' thermal verification plans. This paper describes the JWST Core 2 Test, which is a cryogenic thermal balance test of a full size, high fidelity engineering model of the Observatory's 'Core' area thermal control hardware. The 'Core' area is the key mechanical and cryogenic interface area between all Observatory elements. The 'Core' area thermal control hardware allows for temperature transition of 300K to approximately 50 K by attenuating heat from the room temperature IEC (instrument electronics) and the Spacecraft Bus. Since the flight hardware is not available for test, the Core 2 test uses high fidelity and flight-like reproductions.

Data report for ROSA-IV LSTF gravity-driven safety injection experiment run SB-CL-27

International Nuclear Information System (INIS)

Yonomoto, Taisuke; Saitou, Seishi; Kuroda, Takeshi

1994-03-01

Experimental data are presented for the passive injection test, Run SB-CL-27, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on September 17, 1992. This experiment simulated thermal-hydraulic behavior of a gravity-driven, passive safety injection system during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The injection system consisted of a gravity-driven injection tank, located above the reactor vessel, with connecting lines. The tank was initially filled with water of room temperature at the same pressure as the pressurizer. The connecting lines to the cold leg and to the vessel downcomer were opened at the test initiation. Then, a natural circulation flow developed in the loop which was formed by these lines and the injection tank. The hot water in the cold leg circulated into the upper part of tank and accumulated there causing a significant thermal stratification. This thermal stratification prevented direct-contact condensation of steam from occurring during the subsequent tank drain-down phase. Therefore, no condensation-induced depressurization of the tank, affecting adversely the injection performance, occurred. (author)

SBO simulations for Integrated Passive Safety System (IPSS) using MARS

International Nuclear Information System (INIS)

Kim, Sang Ho; Jeong, Sung Yeop; Chang, Soon Heung

2012-01-01

The current nuclear power plants have lots of active safety systems with some passive safety systems. The safety of current and future nuclear power plants can be enhanced by the application of additional passive safety systems for the ultimate safety. It is helpful to install the passive safety systems on current nuclear power plants without the design change for the licensibility. For solving the problem about the system complexity shown in the Fukushima accidents, the current nuclear power plants are needed to be enhanced by an additional integrated and simplified system. As a previous research, the integrated passive safety system (IPSS) was proposed to solve the safety issues related with the decay heat removal, containment integrity and radiation release. It could be operated by natural phenomena like gravity, natural circulation and pressure difference without AC power. The five main functions of IPSS are: (a) Passive decay heat removal, (b) Passive emergency core cooling, (c) Passive containment cooling, (d) Passive in vessel retention and ex-vessel cooling, and (e) Filtered venting and pressure control. The purpose of this research is to analyze the performances of each function by using MARS code. The simulated accident scenarios were station black out (SBO) and the additional accidents accompanied by SBO

[Application of injection test in confirming the ideal position of esophageal balloon catheter].

Science.gov (United States)

Chen, Han; Xu, Ming; Yang, Yanlin; He, Xuan; Zhou, Jianxin

2017-09-01

To evaluate the safety and feasibility of injection test which is used to locate esophageal balloon catheter. A prospective study was conducted. The patients undergoing invasive mechanical ventilation (MV) admitted to general intensive care unit (ICU) of Beijing Tiantan Hospital Affiliated to Capital Medical University from May 2015 and March 2017 were enrolled. The commercially available esophageal balloon catheter was modified to perform injection test. The catheter was withdrawn step by step and the injection test was repeated until the presence disturbance wave presented, which indicated that the balloon had just entered the esophagus. The position where disturbance wave appears was named 0 cm. End-expiratory occlusions were performed at the positions of +15, +10, +5, 0, -5, -10 and -15 cm, respectively, and the changes of esophageal pressure (Pes) and airway pressures (Paw) were measured in the spontaneous breathing and passive ventilation, and the ratio between the changes (ΔPes/ΔPaw) was calculated. A total of 20 patients were enrolled, of which 15 patients finished both the spontaneous and the passive ventilation parts, and 2 patients finished only the spontaneous part and 3 patients finished only passive part. (1) Disturbance waves could be induced by injection test in all patients. The average depth of disturbance wave in spontaneous breathing was deeper than that in passive ventilation (cm: 42.4±3.8 vs. 41.8±3.3), but there was no significant difference between the two ventilation settings (P = 0.132). No adverse events occurred during the study period. (2) Pes increased with the stepwise withdraw of esophageal catheter, reached the maximal value at +5 cm, and then decreased when the catheter was further withdrawn, no matter in the spontaneous or the passive ventilation. In spontaneous breathing, the ΔPes/ΔPaw was within the ideal range (0.8-1.2) at the positions of 0, -5 and -10 cm. The ΔPes/ΔPaw was closest to unity at the positions of 0 cm (0

Experimental Study and Mathematical Modeling of Asphaltene Deposition Mechanism in Core Samples

Directory of Open Access Journals (Sweden)

Jafari Behbahani T.

2015-11-01

Full Text Available In this work, experimental studies were conducted to determine the effect of asphaltene deposition on the permeability reduction and porosity reduction of carbonate, sandstone and dolomite rock samples using an Iranian bottom hole live oil sample which is close to reservoir conditions, whereas in the majority of previous work, a mixture of recombined oil (a mixture of dead oil and associated gas was injected into a core sample which is far from reservoir conditions. The effect of the oil injection rate on asphaltene deposition and permeability reduction was studied. The experimental results showed that an increase in the oil injection flow rate can result in an increase in asphaltene deposition and permeability reduction. Also, it can be observed that at lower injection flow rates, a monotonic decrease in permeability of the rock samples can be attained upon increasing the injection flow rate, while at higher injection rates, after a decrease in rock permeability, an increasing trend is observed before a steady-state condition can be reached. The experimental results also showed that the rock type can affect the amount of asphaltene deposition, and the asphaltene deposition has different mechanisms in sandstone and carbonate core samples. It can be seen that the adsorption and plugging mechanisms have a more important role in asphaltene deposition in carbonate core samples than sandstone core samples. From the results, it can be observed that the pore volumes of the injected crude oil are higher for sandstone cores compared with the carbonate cores. Also, it can be inferred that three depositional types may take place during the crude oil injection, i.e., continuous deposition for low-permeability cores, slow, steady plugging for high-permeability cores and steady deposition for medium-permeability cores. It can be seen from the experimental results that damage to the core samples was found to increase when the production pressures were

Performance Analysis of Multi Stage Safety Injection Tank

International Nuclear Information System (INIS)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo

2015-01-01

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Performance Analysis of Multi Stage Safety Injection Tank

Energy Technology Data Exchange (ETDEWEB)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Active stabilization of a diode laser injection lock

Energy Technology Data Exchange (ETDEWEB)

Saxberg, Brendan; Plotkin-Swing, Benjamin; Gupta, Subhadeep [Department of Physics, University of Washington, P.O. Box 351560, Seattle, Washington 98195-1560 (United States)

2016-06-15

We report on a device to electronically stabilize the optical injection lock of a semiconductor diode laser. Our technique uses as discriminator the peak height of the laser’s transmission signal on a scanning Fabry-Perot cavity and feeds back to the diode current, thereby maintaining maximum optical power in the injected mode. A two-component feedback algorithm provides constant optimization of the injection lock, keeping it robust to slow thermal drifts and allowing fast recovery from sudden failures such as temporary occlusion of the injection beam. We demonstrate the successful performance of our stabilization method in a diode laser setup at 399 nm used for laser cooling of Yb atoms. The device eases the requirements on passive stabilization and can benefit any diode laser injection lock application, particularly those where several such locks are employed.

Active stabilization of a diode laser injection lock

International Nuclear Information System (INIS)

Saxberg, Brendan; Plotkin-Swing, Benjamin; Gupta, Subhadeep

2016-01-01

We report on a device to electronically stabilize the optical injection lock of a semiconductor diode laser. Our technique uses as discriminator the peak height of the laser’s transmission signal on a scanning Fabry-Perot cavity and feeds back to the diode current, thereby maintaining maximum optical power in the injected mode. A two-component feedback algorithm provides constant optimization of the injection lock, keeping it robust to slow thermal drifts and allowing fast recovery from sudden failures such as temporary occlusion of the injection beam. We demonstrate the successful performance of our stabilization method in a diode laser setup at 399 nm used for laser cooling of Yb atoms. The device eases the requirements on passive stabilization and can benefit any diode laser injection lock application, particularly those where several such locks are employed.

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

Energy Technology Data Exchange (ETDEWEB)

Scheuerer, Martina, E-mail: Martina.Scheuerer@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany); Weis, Johannes, E-mail: Johannes.Weis@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Pressurized thermal shocks are important phenomena for plant life extension and aging. Black-Right-Pointing-Pointer The thermal-hydraulics of PTS have been studied experimentally and numerically. Black-Right-Pointing-Pointer In the Large Scale Test Facility a loss of coolant accident was investigated. Black-Right-Pointing-Pointer CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

How to arrest a core meltdown accident (doing nothing)

International Nuclear Information System (INIS)

Baron, Jorge H.

2000-01-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)

Structural and optical properties of alloyed quaternary CdSeTeS core and CdSeTeS/ZnS core–shell quantum dots

Energy Technology Data Exchange (ETDEWEB)

Adegoke, Oluwasesan, E-mail: adegoke.sesan@mailbox.co.za [Department of Chemistry, Faculty of Natural and Agricultural Sciences, University of Pretoria, Lynnwood Road, Pretoria 0002 (South Africa); Nyokong, Tebello, E-mail: t.nyokong@ru.ac.za [Department of Chemistry, Rhodes University, Grahamstown 6140 (South Africa); Forbes, Patricia B.C., E-mail: patricia.forbes@up.ac.za [Department of Chemistry, Faculty of Natural and Agricultural Sciences, University of Pretoria, Lynnwood Road, Pretoria 0002 (South Africa)

2015-10-05

Highlights: • Alloyed quaternary CdSeTeS core quantum dots (QDs) were synthesized. • Passivation was carried out using a ZnS shell. • Quaternary CdSeTeS core exhibited unique optical properties over CdSeTe/ZnS. • CdSeTeS can be employed as a useful alternative to core/shell QDs. - Abstract: Synthesis of fluorescent alloyed quantum dots (QDs) with unique optical properties suitable for a wide array of chemical, physical and biological applications is of research interest. In this work, highly luminescent and photostable alloyed quaternary CdSeTeS core QDs of two different sizes were fabricated via the organometallic hot-injection synthetic route. Characterization of the nanocrystals were performed using TEM, XRD, UV/vis and fluorescence spectrophotometric techniques. We have demonstrated in this work that the well fabricated alloyed quaternary CdSeTeS core QDs possess unique optical properties that are advantageous over conventional core/shell systems. Formation of the CdSeTeS/ZnS core/shell with the desired optical properties comes with a number of challenges, hence the advantages of the quaternary alloyed core over the core/shell QDs are (i) avoidance of the challenging process of determining the proper shell thickness which can provide the desired optical properties in the core/shell system and (ii) avoidance of the lattice-induced mismatch between the core and the shell material which can either lead to incomplete exciton confinement or dislocation at the core/shell interface.

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

LMFBR core flowering response to an impulse load

International Nuclear Information System (INIS)

Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

1993-01-01

Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

International Nuclear Information System (INIS)

Cho, Dae H.; Page, Richard J.; Abdulla, Sherif H.; Anderson, Mark H.; Klockow, Helge B.; Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb-Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed

Concept of passive safe small reactor for distributed energy supply system

International Nuclear Information System (INIS)

Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

2003-01-01

This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

Effects of upper plenum injection on thermo-hydrodynamic behavior under refill and reflood phases

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Abe, Yutaka; Adachi, Hiromichi; Ohnuki, Akira; Osakabe, Masahiro

1984-12-01

In order to investigate the thermo-hydrodynamic behavior in core under simultaneous ECC water injection into the upper plenum and the intact cold leg during the refill and reflood phases of a PWR-LOCA, Tests S1-SH3 and S1-SH4 were performed by using Slab Core Test Facility (SCTF) with the injection of saturated and 67K subcooled water into the upper plenum, respectively, under the same cold leg injection condition. The following major findings were obtained by examining these test results. (1) Although the core was cooled by the fall back water from the upper plenum into the core during the period of high injection rate into the upper plenum, the core was cooled mainly by the bottom flooding after the BOCREC (Bottom of core recovery). (2) The possible fall back flow rate estimated with a CCFL correlation rapidly decreased after the BOCREC because of the increase of steam generation rate in core. (3) Continuous fall back of subcooled water was not observed even under the condition with large upper plenum injection rate of subcooled water and with steam outflow through the lower plenum into the downcomer. The fall back was intermittently limited by the rapid increase of upward steam flow which was generated in the core due to the evaporation of the fall back water. (4) The rising of liquid level in the lower plenum was suppressed by the pressurization in core due to the evaporation of fall back water before the BOCREC and therefore the beginning of bottom reflood was delayed. Some selected data from Tests S1-SH3 and S1-SH4 are also included in this report. (author)

Natural circulation and stratification in the various passive safety systems of the SWR 1000

International Nuclear Information System (INIS)

Meseth, J.

2002-01-01

In some of the passive safety systems of Siemens' SWR 1000 boiling water reactor (i.e. the emergency condensers and containment cooling condensers), natural circulation is the main effect on both the primary and secondary sides by which optimum system efficiency is achieved. Other passive safety systems of the SWR 1000 require natural circulation on the secondary side only (condensation of steam discharged by the safety and relief valves; cooling of the Reactor Pressure Vessel (RPV) by flooding from the outside in case of core melt), while still other systems require stratification to be effective (i.e. the passive pressure pulse transmitters and steam-driven scram tanks). Complex natural circulation and stratification can take place simultaneously if fluids with different densities are enclosed in a single volume (in a core melt accident, for example, the nitrogen, steam and hydrogen in the containment). Related problems and the solutions thereto planned for the SWR 1000 are reported from the designer's viewpoint. (author)

Passive superconductor: A viable method of controlling magnetization multipoles in the SSC dipole

International Nuclear Information System (INIS)

Green, M.A.

1989-02-01

At injection, the magnetization of the superconductor produces the dominant field error in the SSC dipole magnets. The field generated by magnetization currents in the superconductor is rich in higher symmetric multipoles (normal sextupole, normal decapole, and so on). Pieces of passive superconductor properly located within the bore of the dipole magnet can cancel the higher multipoles generated by the SSC dipole coils. The multipoles generated by the passive superconductor (predominantly sextupole and decapole) are controlled by the angular and radial location of the superconductor, the volume of superconductor, and the size of the superconducting filaments within the passive conductor. This paper will present the tolerances on each of these factors. The paper will show that multipole correction using passive superconductor is in general immune to the effects of temperature and magnetization decay due to flux creep, provided that dipole superconductor and the passive correction superconductor are properly specified. When combined with a lumped correction system, the passive superconductor can be a viable alternative to continuous correction coils within the SSC dipoles. 20 refs., 8 figs., 2 tabs

Passive superconductor a viable method of controlling magnetization multipoles in the SSC dipole

International Nuclear Information System (INIS)

Green, M.A.

1989-01-01

At injection, the magnetization of the superconductor produces the dominant field error in the SSC dipole magnets. The field generated by magnetization currents in the superconductor is rich in higher symmetric multipoles (normal sextupole, normal decapole, and so on). Pieces of passive superconductor properly located within the bore of the dipole magnet can cancel the higher multipoles generated by the SSC dipole coils. The multipoles generated by the passive superconductor (predominantly sextupole and decapole) are controlled by the angular and radial location of the superconductor, the volume of superconductor, and the size of the superconducting filaments within the passive conductor. This paper will present the tolerances on each of these factors. The paper will show that multipole correction using passive superconductor is in general immune to the effects of temperature and magnetization decay due to flux creep, provided that dipole superconductor and the passive correction superconductor are properly specified. When combined with a lumped correction system, the passive superconductor can be a viable alternative to continuous correction coils within the SSC dipoles. 20 refs., 8 figs., 2 tabs

Passive safety features in current and future water cooled reactors

International Nuclear Information System (INIS)

1990-11-01

Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

Metabolic clearance rate and urinary clearance of purified beta-core

International Nuclear Information System (INIS)

Wehmann, R.E.; Blithe, D.L.; Flack, M.R.; Nisula, B.C.

1989-01-01

We injected a highly purified preparation of the beta-core molecule, a fragment of hCG beta excreted in pregnancy urine, into five men and three women to determine its kinetic parameters, MCR, and urinary clearance. The beta-core molecule was distributed in an initial volume [1950 +/- 156 (mean +/- SEM) mL/m2 body surface area] approximately equal to the estimated plasma volume. Its disappearance was multiexponential on a semilogarithmic plot, with a rapid phase t1/2 of 3.5 +/- 0.7 min and a slow phase t1/2 of 22.4 +/- 4.2 min. The transit time (the mean time spent by a molecule of beta-core in transit) was 20.6 +/- 2.1 min. The MCR was 192.0 +/- 8.0 mL/min.m2 body surface area. About 5% of the injected dose of beta-core was excreted into the urine in the first 30 min after injection, and low levels of excretion persisted for up to 7 days. The urinary clearance rate of beta-core was 13.7 +/- 1.4 mL/min.m2, accounting for about 8% of the elimination of beta-core from the plasma. The beta-core immunoreactivity in serum and urine was characterized by gel filtration and three independent RIA systems to show that its properties were indistinguishable from those of the injected beta-core. Serum levels of beta-core in pregnant women were less than 0.2 ng/mL, while the amounts excreted in their urine were as much as 5 mg/day. Based on these clearance parameters of beta-core in normal subjects, less than 0.2% of the beta-core excreted in pregnancy urine is derived by urinary clearance of plasma beta-core. Therefore, more than 99% of the beta-core excreted in pregnancy urine is derived from beta-core in a compartment separate from plasma. In particular, these data indicate that there is relatively little placental secretion of beta-core into plasma and that placental secretion does not account for the vast majority of beta-core in pregnancy urine

Passive cooling containment study

International Nuclear Information System (INIS)

Shin, J.J.; Iotti, R.C.; Wright, R.F.

1993-01-01

Pressure and temperature transients of nuclear reactor containment following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally for the full scale NPR (New Production Reactor). All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified CONTEMPT code CONTEMPT-NPR. In order to verify proper heat transfer coefficient, temperature, heat flux, and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and half foot (1.7m) diameter outer cylinder. Comparison of CONTEMPT-NPR and WGOTHIC was done for reduced scale NPR

Emergency core cooling device

International Nuclear Information System (INIS)

Suzaki, Kiyoshi; Inoue, Akihiro.

1979-01-01

Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

Plasma core reactor applications

International Nuclear Information System (INIS)

Latham, T.S.; Rodgers, R.J.

1976-01-01

Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

Process Optimization for Injection Moulding of Passive Microwave Components

DEFF Research Database (Denmark)

Scholz, Steffen G.; Mueller, Tobias; Santos Machado, Leonardo

2016-01-01

The demand for micro components has increased during the last decade following the overall trend towards miniaturization. Injection moulding is the favoured technique for the mass manufacturing of micro components or larger parts with micro-structured areas due to its ability to cost effectively ...... algorithm for modelling, the influence of different moulding parameters on the final part quality was assessed. Firstly a process model and secondly a quality model has been calculated. The results shows that part quality can be controlled by monitoring characteristic numbers....

Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

2004-06-01

The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

The Development of Computer Code for Safety Injection Tank (SIT) with Fluidic Device(FD) Blowdown Test

International Nuclear Information System (INIS)

Lee, Joo Hee; Kim, Tae Han; Choi, Hae Yun; Lee, Kwang Won; Chung, Chang Kyu

2007-01-01

Safety Injection Tanks (SITs) with the Fluidic Device (FD) of APR1400 provides a means of rapid reflooding of the core following a large break Loss Of Coolant Accident (LOCA), and keeping it covered until flow from the Safety Injection Pump (SIP) becomes available. A passive FD can provide two operation stages of a safety water injection into the RCS and allow more effective use of borated water in case of LOCA. Once a large break LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced to a lower flow rate. The conventional computer code 'TURTLE' used to simulate the blowdown of OPR1000 SIT can not be directly applied to simulate a blowdown process of the SIT with FD. A new computer code is needed to be developed for the blowdown test evaluation of the APR1400 SIT with FD. Korea Power Engineering Company (KOPEC) has developed a new computer code to analyze the characteristics of the SIT with FD and validated the code through the comparison of the calculation results with the test results obtained by Ulchin 5 and 6 units pre-operational test and VAlve Performance Evaluation Rig (VAPER) tests performed by The Korea Atomic Energy Research Institute (KAERI)

Sum rules and constraints on passive systems

International Nuclear Information System (INIS)

Bernland, A; Gustafsson, M; Luger, A

2011-01-01

A passive system is one that cannot produce energy, a property that naturally poses constraints on the system. A system in convolution form is fully described by its transfer function, and the class of Herglotz functions, holomorphic functions mapping the open upper half-plane to the closed upper half-plane, is closely related to the transfer functions of passive systems. Following a well-known representation theorem, Herglotz functions can be represented by means of positive measures on the real line. This fact is exploited in this paper in order to rigorously prove a set of integral identities for Herglotz functions that relate weighted integrals of the function to its asymptotic expansions at the origin and infinity. The integral identities are the core of a general approach introduced here to derive sum rules and physical limitations on various passive physical systems. Although similar approaches have previously been applied to a wide range of specific applications, this paper is the first to deliver a general procedure together with the necessary proofs. This procedure is described thoroughly and exemplified with examples from electromagnetic theory.

Effects of Ambient Temperature and Forced-air Warming on Intraoperative Core Temperature: A Factorial Randomized Trial.

Science.gov (United States)

Pei, Lijian; Huang, Yuguang; Xu, Yiyao; Zheng, Yongchang; Sang, Xinting; Zhou, Xiaoyun; Li, Shanqing; Mao, Guangmei; Mascha, Edward J; Sessler, Daniel I

2018-05-01

The effect of ambient temperature, with and without active warming, on intraoperative core temperature remains poorly characterized. The authors determined the effect of ambient temperature on core temperature changes with and without forced-air warming. In this unblinded three-by-two factorial trial, 292 adults were randomized to ambient temperatures 19°, 21°, or 23°C, and to passive insulation or forced-air warming. The primary outcome was core temperature change between 1 and 3 h after induction. Linear mixed-effects models assessed the effects of ambient temperature, warming method, and their interaction. A 1°C increase in ambient temperature attenuated the negative slope of core temperature change 1 to 3 h after anesthesia induction by 0.03 (98.3% CI, 0.01 to 0.06) °Ccore/(h°Cambient) (P ambient temperature with passive insulation, but was unaffected by ambient temperature during forced-air warming (0.02 [98.3% CI, -0.04 to 0.09] °Ccore/°Cambient; P = 0.40). After an average of 3.4 h of surgery, core temperature was 36.3° ± 0.5°C in each of the forced-air groups, and ranged from 35.6° to 36.1°C in passively insulated patients. Ambient intraoperative temperature has a negligible effect on core temperature when patients are warmed with forced air. The effect is larger when patients are passively insulated, but the magnitude remains small. Ambient temperature can thus be set to comfortable levels for staff in patients who are actively warmed.

TEPSS - Technology Enhancement for Passive Safety Systems

International Nuclear Information System (INIS)

Hart, J.; Slegers, W.J.M.; Boer, S.L. de; Huggenberger, M.; Lopez Jimenez, J.; Munoz-Cabo Gonzalez, J.L.; Reventos Puigjaner, F.

2000-01-01

The objective of the TEPSS project was to make significant additions to the technology base of the European Simplified Boiling Water Reactor (ESBWR). The project focused on mixing and stratification phenomena in large water pools, passive decay heat removal from containments, and effects of aerosol deposition inside a passive heat exchanger. The PSI experimental facility LINX (Large-scale Investigation of Natural Circulation and Mixing) has been used to investigate venting of steam and steam-noncondensable gas mixtures into water pools. The test revealed that no significant steam bypass could be detected when injecting a mixture of steam or air and that mixing was very efficient. In addition to the tests, 3-D numerical computations and initial model development have been performed to study the behaviour of bubble plumes in water pools. The major part of the TEPSS project studied selective aspects of the response technology of modem pressure-suppression type containment designs and of passive-type decay heat removal systems. The work included an experimental phase using the large-scale experimental facility PANDA (Passive Nachwaermeabfuhr und Druckabbau), operated by PSI, where eight experiments successfully have been executed to test the performance of the ESBWR containment configuration. The PANDA tests have been analysed successfully using thermalhydraulic system analysis codes and 3-D CFD codes. The AIDA (Aerosol Impaction and Deposition Analysis) experimental facility of PSI has been used to investigate the degradation of passive decay heat removal due to fission product aerosols deposited on the inside surfaces of the PCC (Passive Containment Cooler) heat exchanger tubes. The one test performed revealed that the degradation of the heat transfer in the PCC tubes due to the deposition of aerosols reached about 20%. The test has been analysed using the MELCOR severe accident analysis code. (author)

Control of Surge in Centrifugal Compressor by Using a Nozzle Injection System: Universality in Optimal Position of Injection Nozzle

Directory of Open Access Journals (Sweden)

Toshiyuki Hirano

2012-01-01

Full Text Available The passive control method for surge and rotating stall in centrifugal compressors by using a nozzle injection system was proposed to extend the stable operating range to the low flow rate. A part of the flow at the scroll outlet of a compressor was recirculated to an injection nozzle installed on the inner wall of the suction pipe of the compressor through the bypass pipe and injected to the impeller inlet. Two types of compressors were tested at the rotational speeds of 50,000 rpm and 60,000 rpm with the parameter of the circumferential position of the injection nozzle. The present experimental results revealed that the optimum circumferential position, which most effectively reduced the flow rate for the surge inception, existed at the opposite side of the tongue of the scroll against the rotational axis and did not depend on the compressor system and the rotational speeds.

Preparation of Fe(3)O(4)@C@CNC multifunctional magnetic core/shell nanoparticles and their application in a signal-type flow-injection photoluminescence immunosensor.

Science.gov (United States)

Chu, Chengchao; Li, Meng; Li, Long; Ge, Shenguang; Ge, Lei; Yu, Jinghua; Yan, Mei; Song, Xianrang

2013-11-01

We describe here the preparation of carbon-coated Fe3O4 magnetic nanoparticles that were further fabricated into multifunctional core/shell nanoparticles (Fe3O4@C@CNCs) through a layer-by-layer self-assembly process of carbon nanocrystals (CNCs). The nanoparticles were applied in a photoluminescence (PL) immunosensor to detect the carcinoembryonic antigen (CEA), and CEA primary antibody was immobilized onto the surface of the nanoparticles. In addition, CEA secondary antibody and glucose oxidase were covalently bonded to silica nanoparticles. After stepwise immunoreactions, the immunoreagent was injected into the PL cell using a flow-injection PL system. When glucose was injected, hydrogen peroxide was obtained because of glucose oxidase catalysis and quenched the PL of the Fe3O4@C@CNC nanoparticles. The here proposed PL immunosensor allowed us to determine CEA concentrations in the 0.005–50 ng·mL-1 concentration range, with a detection limit of 1.8 pg·mL-1.

Numerical investigation of passive heat removal system via steam generator in VVER 1200

International Nuclear Information System (INIS)

Dinh Anh Tuan; Duong Thanh Tung; Tran Chi Thanh; Nguyen Van Thai

2015-01-01

Passive heat removal system (PHRS) via Steam Generator is an important part in VVER design. In case of Design Basic Accidents such as blackout, failure of feed water supply to steam generator or coolant leakage with failure of emergency core cooling at high pressure. PHRS is designed to remove the residual heat from reactor core through steam generator to heat exchanger which is placed outside reactor vessel. In order to evaluate the passive system, a numerical investigation using a CFD code is performed. However, PHRS has complex geometry for using CFD simulation. Thus, RELAP5 is applied to provide the wall heat flux of tube in the heat exchanger tank. The natural convection in the heat exchanger tank is investigated in this report. Numerical results show temperature and velocity distribution in the heat exchanger tank are calculated with different wall heat flux corresponding to various transient conditions. The calculated results contribute to the capacity analysis of passive heat removal system and giving valuable information for safe operation of VVER 1200. (author)

Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

Correction of magnetization sextupole and decapole in a 5 centimeter bore SSC dipole using passive superconductor

International Nuclear Information System (INIS)

Green, M.A.

1991-05-01

Higher multipoles due to magnetization of the superconductor in four and five centimeter bore Superconducting Super Collider (SSC) superconducting dipole magnets have been observed. The use of passive superconductor to correct out the magnetization sextupole has been demonstrated on two dipoles built by the Lawrence Berkeley Laboratory (LBL). This reports shows how passive correction can be applied to the five centimeter SSC dipoles to remove sextupole and decapole caused by magnetization of the dipole superconductor. Two passive superconductor corrector options will be presented. The change in magnetization sextupole and decapole due to flux creep decay of the superconductor during injection can be partially compensated for using the passive superconductor. 9 refs; 5 figs

Sulfur passivation and contact methods for GaAs nanowire solar cells

International Nuclear Information System (INIS)

Tajik, N; Peng, Z; Kuyanov, P; LaPierre, R R

2011-01-01

The effect of sulfur passivation on core-shell p-n junction GaAs nanowire (NW) solar cells has been investigated. Devices of two types were investigated, consisting of indium tin oxide contact dots or opaque Au finger electrodes. Lateral carrier transport from the NWs to the contact fingers was achieved via a p-doped GaAs surface conduction layer. NWs between the opaque contact fingers had sidewall surfaces exposed for passivation by sulfur. The relative cell efficiency increased by 19% upon passivation. The contribution of the thin film grown between the NWs to the total cell efficiency was estimated by removing the NWs using a sonication procedure. Mechanisms of carrier transport and photovoltaic effects are discussed on the basis of spatially resolved laser scanning measurements.

Passive sampling and analyses of common dissolved fixed gases in groundwater

International Nuclear Information System (INIS)

Spalding, Brian Patrick; Watson, David B.

2008-01-01

An in situ passive sampler and gas chromatographic protocol for analysis of the major and several minor fixed gases in groundwater was developed. A gas-tight syringe, mated to a short length of silicone tubing, was equilibrated with dissolved gases in groundwater by immersing in monitoring wells and was used to transport and to inject a 0.5 mL gas sample into a gas chromatograph. Using Ar carrier gas, a HaySep DB porous polymer phase, and sequential thermal conductivity and reductive gas detectors allowed good sensitivity for He, Ne, H2, N2, O2, CO, CH4, CO2, and N2O. Within 4 days of immersion in groundwater, samplers initially filled with either He or air attained the same and constant gas composition at an Oak Ridge, Tennessee, site heavily impacted by uranium, acidity, and nitrate. Between June 2006 and July 2007, 12 permanent groundwater wells were used to test the passive samplers in groundwater contaminated by a group of four closed radioactive wastewater seepage ponds; over a thousand passive gas samples from these wells averaged 56% CO2, 32.4% N2, 2.5% O2, 2.5% N2O, 0.20% CH4, 0.096% H2, and 0.023% CO with an average recovery of 95 14% of the injected gas volume

ASTRID core: Design objectives, design approach, and R&D in support

International Nuclear Information System (INIS)

Mignot, G.; Devictor, N.

2012-01-01

ASTRID core design is mainly guided by safety objectives: 1. Prevention of the core meltdown accident: To prevent meltdown accidents: - by a natural behavior of the core and the reactor (no actuation of the two shutdown systems); - with adding passive complementary systems if natural behavior is not sufficient for some transient cases. 2. Mitigation of the fusion accident: To garantee that core fusion accidents don’t lead to significant mechanical energy release, whatever initiator event: - by a natural core behavior; - with adding specific mitigation dispositions in case of natural behavior is not suffficient

AA injection kicker in its tank

CERN Multimedia

CERN PhotoLab

1980-01-01

For single-turn injection of the antiprotons, a septum at the end of the injection line made the beam parallel to the injection orbit, and a quarter of a betatron-wavelength downstream a fast kicker corrected the angle. Kicker type: lumped delay line. PFN voltage 56 kV. Bending angle 7.5 mrad; kick-strength 0.9 Tm; fall-time 95%-5% in 150 ns. The injection orbit is to the left, the stack orbit to the far right. A fast shutter near the central orbit had to be closed before the kicker fired, so as to protect the stack core from being shaken by the kicker's fringe field. The shutter is shown in closed position.

CMT scaling analysis and distortion evaluation in passive integral test facility

International Nuclear Information System (INIS)

Deng Chengcheng; Qin Benke; Wang Han; Chang Huajian

2013-01-01

Core makeup tank (CMT) is the crucial device of AP1000 passive core cooling system, and reasonable scaling analysis of CMT plays a key role in the design of passive integral test facilities. H2TS method was used to perform scaling analysis for both circulating mode and draining mode of CMT. And then, the similarity criteria for CMT important processes were applied in the CMT scaling design of the ACME (advanced core-cooling mechanism experiment) facility now being built in China. Furthermore, the scaling distortion results of CMT characteristic Ⅱ groups of ACME were calculated. At last, the reason of scaling distortion was analyzed and the distortion evaluation was conducted for ACME facility. The dominant processes of CMT circulating mode can be adequately simulated in the ACME facility, but the steam condensation process during CMT draining is not well preserved because the excessive CMT mass leads to more energy to be absorbed by cold metal. However, comprehensive analysis indicates that the ACME facility with high-pressure simulation scheme is able to properly represent CMT's important phenomena and processes of prototype nuclear plant. (authors)

Design of the injection channel magnets for the K1200 cyclotron

International Nuclear Information System (INIS)

Wu, X.Y.; Lawton, D.; Marti, F.; Zeller, A.F.

2001-01-01

The beam from the K500 cyclotron is injected radially into the median plane of the K1200 cyclotron where it is stripped inside a dee. Along the injection path, the injected beam traverses large magnetic gradients under the K1200 superconducting coil and on the edge of the hill. To match the injected beam to the eigen-ellipse for the accelerated beam, two magnetic elements are included in the injection channel in the yoke of the K1200 cyclotron. One of them is an active element and the other a conical focusing passive element. This last element must move in the horizontal plane to match the different beam paths. We describe in the paper the beam dynamics calculations as well as the mechanical design and construction of these magnets

Assessment of GOTHIC and TRACE codes against selected PANDA experiments on a Passive Containment Condenser

Energy Technology Data Exchange (ETDEWEB)

Papini, Davide, E-mail: davide.papini@psi.ch; Adamsson, Carl; Andreani, Michele; Prasser, Horst-Michael

2014-10-15

Highlights: • Code comparison on the performance of a Passive Containment Condenser. • Simulation of separate effect tests with pure steam and non-condensable gases. • Role of the secondary side and accuracy of pool boiling models are discussed. • GOTHIC and TRACE predict the experimental performance with slight underestimation. • Recirculatory flow pattern with injection of light non-condensable gas is inferred. - Abstract: Typical passive safety systems for ALWRs (Advanced Light Water Reactors) rely on the condensation of steam to remove the decay heat from the core or the containment. In the present paper the three-dimensional containment code GOTHIC and the one-dimensional system code TRACE are compared on the calculation of a variety of phenomena characterizing the response of a passive condenser submerged in a boiling pool. The investigation addresses the conditions of interest for the Passive Containment Cooling System (PCCS) proposed for the ESBWR (Economic Simplified Boiling Water Reactor). The analysis of selected separate effect tests carried out on a PCC (Passive Containment Condenser) unit in the PANDA large-scale thermal-hydraulic facility is presented to assess the code predictions. Both pure steam conditions (operating pressure of 3 bar, 6 bar and 9 bar) and the effect on the condensation heat transfer of non-condensable gases heavier than steam (air) and lighter than steam (helium) are considered. The role of the secondary side (pool side) heat transfer on the condenser performance is examined too. In general, this study shows that both the GOTHIC and TRACE codes are able to reasonably predict the heat transfer capability of the PCC as well as the influence of non-condensable gas on the system. A slight underestimation of the condenser performance is obtained with both codes. For those tests where the experimental and simulated efficiencies agree better the possibility of compensating errors among different parts of the heat transfer

All-VCSEL Transmitters With Remote Optical Injection for WDM-OFDM-PON

DEFF Research Database (Denmark)

Deng, Lei; Zhao, Ying; Pang, Xiaodan

2014-01-01

We report on a novel scheme that uses vertical cavity surface emitting lasers (VCSELs) and remote optical injection technique in the hybrid wavelength division multiplexing orthogonal frequency division multiplexing (OFDM) passive optical network. In the proposed scheme, 1.55-$\\mu{\\rm m}$ VCSELs ...

Passive safety features for next generation CANDU power plants

International Nuclear Information System (INIS)

Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

1989-01-01

CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

Passive residual energy utilization system in thermal cycles on water-cooled power reactors

International Nuclear Information System (INIS)

Placco, Guilherme M.; Guimaraes, Lamartine N.F.; Santos, Rubens S. dos

2013-01-01

This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

Study of passive residual heat removal system of a modular small PWR reactor

International Nuclear Information System (INIS)

Araujo, Nathália N.; Su, Jian

2017-01-01

This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

Design of the core support and restraint structures for FFTF and CRBRP

International Nuclear Information System (INIS)

Sutton, H.G.; Rylatt, J.A.

1977-12-01

This paper presents and compares the design and fabrication of the FFTF and CRBRP reactor structures which support and restrain the reactor core assemblies. The fabrication of the core support structure (CSS) for the FFTF reactor was completed October 1972 and this paper discusses how the fabrication problems encountered with the FFTF were avoided in the subsequent design of the CRBR CSS. The radial core restraint structure of the FFTF was designed and fabricated such that an active system could replace the present passive system which is segmented and relies on the CSS core barrel for total structure integrity to maintain core geometry. The CRBR core restraint structure is designed for passive restraint only, and this paper discusses how the combined strengths of the restraint structure former rings and the CSS core barrel are utilized to maintain core geometry. Whereas the CSS for the FFTF interfaces directly with the reactor core assemblies, the CRBR CSS does not. A comparison is made on how intermediate structures in CRBR (inlet modules) provide the necessary design interfaces for supporting and providing flow distribution to the reactor core assemblies. A discussion is given on how the CRBR CSS satisfied the design requirements of the Equipment Specification, including thermal transient, dynamic and seismic loadings, and results of flow distribution testing that supported the CRBR design effort. The approach taken to simplify fabrication of the CRBR components, and a novel 20 inch deep narrow gap weld joint in the CSS are described

Evaluation of Core Loss in Magnetic Materials Employed in Utility Grid AC Filters

DEFF Research Database (Denmark)

Beres, Remus Narcis; Wang, Xiongfei; Blaabjerg, Frede

2016-01-01

magnetic materials adopted in utility grid ac filters have been investigated and measured for both sinusoidal and rectangular excitation, with and without dc bias condition. The core loss information can ensure cost effective passive filter designs and may avoid trial-error design procedures of the passive......Inductive components play an important role in filtering the switching harmonics related to the pulse width modulation in voltage source converters. Particularly, the filter reactor on the converter side of the filter is subjected to rectangular excitation which may lead to significant losses...... in the core, depending on the magnetic material of choice and current ripple specifications. Additionally, shunt or series reactors that exists in LCL or trap filters and which are subjected to sinusoidal excitations have different specifications and requirements. Therefore, the core losses of different...

From dodewaard to a modern economic passive plant-ESBWR

International Nuclear Information System (INIS)

Gonzalez Lopez, A.; Arnold, H.; Yadigaroglu, G.; Stoop, P.M.; Rao, A.

1997-01-01

For over 25 years the Dodewaard nuclear plant has produced electricity with one of the highest reliability of any power plant in the world. Almost 10 years ago when some designers looked at the features to incorporate in a modern mid-size BWR design, they chose some of the key features of Dodewaard natural circulation and passive safety. Over the years this design evolved into a 670 SBWR design that was developed by an international team and consisted of natural circulation and passive safety injection and decay heat removal. Since the passive decay heat removal was a major new technology area, an extensive test program was developed and conducted utilizing newly constructed large scale integrated system test facility at the Paul Scherrer Institute, Switzerland. The development of any modern reactor is time consuming and expensive, hence, the design and technology was done as part of an international team effort. This paper provides an overview of the international design and technology effort and discusses how the development costs were minimized through cooperation. (Author)

Passive autocatalytic recombiners for combustible gas control in advanced light water reactors

International Nuclear Information System (INIS)

Wolff, U.; Sliter, G.

2004-01-01

A key aspect of the worldwide effort to develop advanced nuclear power plants is designing to address severe accident phenomena, including the generation of hydrogen during core melt progression (metal-water and core-concrete reactions). This design work not only resolves safety concerns with hydrogen, but also supports the development of a technical basis for simplification of off-site emergency planning. The dominant challenge to any emergency planning approach is a large, early containment failure due to pressure excursions. Among the potential contributors to large and rapid increases in containment pressure is hydrogen combustion. The more improbable a containment-threatening combustion becomes, the more appropriate the argument for significant emergency planning simplification. As discussed in this paper, catalytic recombiners provide a means to passively and reliably limit hydrogen combustion to a continuous oxidation process with virtually no potential for containment failure in passive advanced light water reactors (ALWRs). (author)

Study on thermal-hydraulic phenomena identification of passive heat removal facilities

International Nuclear Information System (INIS)

Park, J. Y.

2011-01-01

Recently, passive heat removal facilities have been integral features of new generation or future reactor designs worldwide. This is because the passive heat removal facilities depending on a natural force such as buoyancy can give much higher operational reliability compared to active heat removal facilities depending on pumped fluid flow and as a result they can decrease core damage frequency of a nuclear power plant drastically ever achievable before. Keeping pace with this global trend, SMART and APR+ reactors also have introduced passive heat removal features such as a passive residual heat removal system (PRHRS) and a passive auxiliary feed water system (PAFS) in their designs. Since many thermal-hydraulic (T-H) phenomena including steam condensation are involved during operation of the passive heat removal facilities, they ought to be properly simulated by T-H codes such as MARS-KS and RELAP5 in order to guarantee reliable safety analysis by these codes. Unfortunately, however, these T-H codes are not well validated with respect to phenomena related to passive heat removal mechanism because previous focus on these codes validation was mainly on the LB LOCA and resulting phenomena. To resolve this gap, Korea Institute of Nuclear Safety has initiated a research program on the development of safety analysis technology for passive heat removal facilities. The main target of this program is PRHRS and PAFS in SMART and APR+ reactors and through this program, validation of capability of existing T-H codes and improvement of codes regarding passive facilities analysis are to be sought. In part of this research, T-H phenomena important to passive heat removal facilities (PRHRS and PAFS) are investigated in the present study

Core Sulphate-Reducing Microorganisms in Metal-Removing Semi-Passive Biochemical Reactors and the Co-Occurrence of Methanogens

Directory of Open Access Journals (Sweden)

Maryam Rezadehbashi

2018-02-01

Full Text Available Biochemical reactors (BCRs based on the stimulation of sulphate-reducing microorganisms (SRM are emerging semi-passive remediation technologies for treatment of mine-influenced water. Their successful removal of metals and sulphate has been proven at the pilot-scale, but little is known about the types of SRM that grow in these systems and whether they are diverse or restricted to particular phylogenetic or taxonomic groups. A phylogenetic study of four established pilot-scale BCRs on three different mine sites compared the diversity of SRM growing in them. The mine sites were geographically distant from each other, nevertheless the BCRs selected for similar SRM types. Clostridia SRM related to Desulfosporosinus spp. known to be tolerant to high concentrations of copper were members of the core microbial community. Members of the SRM family Desulfobacteraceae were dominant, particularly those related to Desulfatirhabdium butyrativorans. Methanogens were dominant archaea and possibly were present at higher relative abundances than SRM in some BCRs. Both hydrogenotrophic and acetoclastic types were present. There were no strong negative or positive co-occurrence correlations of methanogen and SRM taxa. Knowing which SRM inhabit successfully operating BCRs allows practitioners to target these phylogenetic groups when selecting inoculum for future operations.

Evaluation report on SCTF Core-III Test S3-22

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; Akimoto, Hajime; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-07-01

Two tests (Tests S3-20 and S3-22) were conducted with JAERI's Slab Core Test Facility (SCTF) Core-III in order to investigate water break-through and core cooling behaviors under the intermittent ECC water delivery from the hot legs to one location in the upper plenum and the alternate ECC water delivery to two locations in the upper plenum during reflooding, respectively. This report presents an analysis on Test S3-22 (the alternate case). Subcooled ECC water was injected alternately just above the upper core support plate above Bundles 7 and 8 and Bundles 3 and 4. The total injection rate from both injection ports was the same as that in SCTF Test S3-20 and Test S3-13. Analyzing the test data together with those of Tests S3-13 and S3-20 the following has been found: (1) Alternate break-through occurred immediately corresponding to the alternate ECC water injection except for one period, during which no break-through was observed. However, there observed a difference in break-through behavior that break-through was strong above the low power region, whereas weak above the high power region. (2) Although its break-through behavior was different, nearly the same core cooling as in the continuous or intermittent ECC water delivery case was observed except for the period around quench. (3) Around quench time, degraded core cooling comparing to the continuous or intermittent ECC water delivery case was observed. That is, quench time at the midplane level of the present test was 35 s later than in the continuous case. This is considered to result from decrease in core water inventory caused by water sealing at the cross-over leg. (J.P.N.)

Evaluation report on SCTF Core-III test S3-06

International Nuclear Information System (INIS)

Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Adachi, Hiromichi; Murao, Yoshio; Minato, Akihiko; Sakaki, Isao.

1988-10-01

In order to investigate the effect of radial power distribution on the thermal-hydraulic characteristics during the reflood phase of a PWR-LOCA with a combined injection type ECCS, a core cooling separate effect test S3-06 and a combined injection test S3-16-Phase 2 were performed using the Slab Core Test Facility (SCTF) Core-III. The radial power distributions in these two tests simulated a reference distribution for a PWR with a combined injection type ECCS and a steep distribution for a PWR with a cold leg injection type ECCS, respectively. Under the radial power distribution of a PWR with a combined injection type ECCS, the radial power distribution had little effect on the thermal-hydraulic behavior in the two-phase up-flow region due to the approximately flat power distribution in this region (power ratio = 1.04 ∼ 1.08). The overall fluid behavior in the pressure vessel was also little affected by the radial power distribution. On the other hand, under the steep radial power distribution (peak power ratio = 1.36), the degree of heat transfer enhancement in high power bundles in the two-phase up-flow region was dominated by the bundlewise radial power ratio as in the case of a PWR with a cold leg injection type ECCS. (author)

Relationship between misonidazole toxicity and core temperature in C3H mice

International Nuclear Information System (INIS)

Gomer, C.J.; Johnson, R.J.

1979-01-01

A single intraperitoneal injection of the radiation sensitizer misonidazole at doses greater than 0.5 mg/g was found to produce a transient hypothermic response in C3H mice. An increase in the acute toxicity of this drug was demonstrated when the animal core temperature was maintained at a normal 35 to 37 0 C by placing the mice in a warmed environment immediately following injection of the drug. The LD/sub 50/3 days/ dose of misonidazole was determined to be 1.48 mg/g for mice allowed to become hypothermic following injection but 0.77 mg/g for mice maintained at a normal core temperature following injection

Accident tolerant high-pressure helium injection system concept for light water reactors

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

Passive safety testing at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Lucoff, D.M.

1989-01-01

During 1986, the Fast Flux Test Facility (FFTF) conducted several tests designed to improve the understanding of the passive safety characteristics of an oxide-fueled liquid-metal reactor (LMR). Static and dynamic tests were performed over a broad range of power, flow, and temperature conditions that extended beyond those for normal operation. Key results of these tests are presented. Stable operation at low power with natural circulation cooling was demonstrated. A passive safety enhancement feature, the gas expansion module (GEM) was developed specifically to offset the large amount of cooldown reactivity that needs to be controlled in an oxide-fueled LMR undergoing an unprotected loss-of-flow accident. Nine GEMs were built and successfully tested in FFTF. With the reactor at 50% power (200 MW (thermal)), the main coolant pumps were turned off and the normal control rod scram response was inhibited. The GEMs and inherent core reactivity feedback mechanisms took the core subcritical with a modest peak coolant temperature transient that reached 85 degrees C above the pretransient value and always maintained a >400 degrees C margin to the sodium boiling point (910 degrees C)

Modelling of passive heating for replication of sub-micron patterns in optical disk substrates

Energy Technology Data Exchange (ETDEWEB)

Kim, Youngmin; Bae, Jaecheol; Kim, Hongmin; Kang, Shinill [School of Mechanical Engineering, Yonsei University, 134 Shinchon-dong, Seodaemoon-ku, Seoul (Korea, Republic of)

2004-05-07

The transcribability of pit or land groove structures in replicating an optical disk substrate greatly affects the performance of a high-density optical disk. However, a solidified layer generated during the polymer filling worsens transcribability because the solidified layer prevents the polymer melt from filling the sub-micron patterns. Therefore, the development of the solidified layer during the filling stage of injection moulding must be delayed. For this delay, passive heating through an insulation layer has been used. In the present study, to examine the development of the solidified layer, delayed by passive heating, the flow of the polymer melt with passive heating was analysed. Passive heating delayed markedly the development of the solidified layer, reduced the viscosity of the polymer melt, and increased the fluidity of the polymer melt in the vicinity of the stamper surface with the sub-micron patterns. As a result, we predict that passive heating can improve the transcribability of an optical disk substrate. To verify our prediction, we fabricated an optical disk substrate by using passive heating of a mould and measured the transcribability of an optical disk substrate.

Modelling of passive heating for replication of sub-micron patterns in optical disk substrates

International Nuclear Information System (INIS)

Kim, Youngmin; Bae, Jaecheol; Kim, Hongmin; Kang, Shinill

2004-01-01

The transcribability of pit or land groove structures in replicating an optical disk substrate greatly affects the performance of a high-density optical disk. However, a solidified layer generated during the polymer filling worsens transcribability because the solidified layer prevents the polymer melt from filling the sub-micron patterns. Therefore, the development of the solidified layer during the filling stage of injection moulding must be delayed. For this delay, passive heating through an insulation layer has been used. In the present study, to examine the development of the solidified layer, delayed by passive heating, the flow of the polymer melt with passive heating was analysed. Passive heating delayed markedly the development of the solidified layer, reduced the viscosity of the polymer melt, and increased the fluidity of the polymer melt in the vicinity of the stamper surface with the sub-micron patterns. As a result, we predict that passive heating can improve the transcribability of an optical disk substrate. To verify our prediction, we fabricated an optical disk substrate by using passive heating of a mould and measured the transcribability of an optical disk substrate

Passive safety design characteristics of the KALIMER-600 burner reactor

International Nuclear Information System (INIS)

Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

2009-01-01

The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

Effects of Starch on Properties of Alumina-based Ceramic Cores

Directory of Open Access Journals (Sweden)

LI Fengguang

2016-12-01

Full Text Available In order to improve the poor leachability of alumina-based ceramic cores, different amount of starch was added to the specimens as pore former. Alumina-based ceramic cores were prepared by hot injection technology using corundum powder as base material, paraffin wax and beeswax as plasticizer, silica powder and magnesium oxide powder as mineralizing agent, wherein the parameters of the hot injection process were as follows:temperature of the slurry was 90℃, hot injection pressure was 0.5 MPa and holding time was 25 s. The effects of starch content on the properties of alumina-based ceramic cores were studied and discussed. The results indicate that during sintering period, the loss of starch in the specimens makes porosity of the alumina-based ceramic cores increase. When starch content increases, the room-temperature flexural strength of the ceramic cores reduces and the apparent porosity increases; the volatile solvent increases and the bulk density decreases. After being sintered at 1560℃ for 2.5 h, room-temperature flexural strength of the alumina-based ceramic cores with starch content of 8%(mass fraction is 24.8 MPa, apparent porosity is 47.98% when the volatile solvent is 1.92 g/h and bulk density is 1.88 g/cm3, the complex properties are optimal.

Measurements of waste tank passive ventilation rates using tracer gases

International Nuclear Information System (INIS)

Huckaby, J.L.; Olsen, K.B.; Sklarew, D.S.; Evans, J.C.; Remund, K.M.

1997-09-01

This report presents the results of ventilation rate studies of eight passively ventilated high-level radioactive waste tanks using tracer gases. Head space ventilation rates were determined for Tanks A-101, AX-102, AX-103, BY-105, C-107, S-102, U-103, and U-105 using sulfur hexafluoride (SF 6 ) and/or helium (He) as tracer gases. Passive ventilation rates are needed for the resolution of several key safety issues. These safety issues are associated with the rates of flammable gas production and ventilation, the rates at which organic salt-nitrate salt mixtures dry out, and the estimation of organic solvent waste surface areas. This tracer gas study involves injecting a tracer gas into the tank headspace and measuring its concentration at different times to establish the rate at which the tracer is removed by ventilation. Tracer gas injection and sample collection were performed by SGN Eurisys Service Corporation and/or Lockheed Martin Hanford Corporation, Characterization Project Operations. Headspace samples were analyzed for He and SF 6 by Pacific Northwest National Laboratory (PNNL). The tracer gas method was first demonstrated on Tank S-102. Tests were conducted on Tank S-102 to verify that the tracer gas was uniformly distributed throughout the tank headspace before baseline samples were collected, and that mixing was sufficiently vigorous to maintain an approximately uniform distribution of tracer gas in the headspace during the course of the study. Headspace samples, collected from a location about 4 in away from the injection point and 15, 30, and 60 minutes after the injection of He and SF 6 , indicated that both tracer gases were rapidly mixed. The samples were found to have the same concentration of tracer gases after 1 hour as after 24 hours, suggesting that mixing of the tracer gas was essentially complete within 1 hour

Fundamental studies of passivity and passivity breakdown

International Nuclear Information System (INIS)

Macdonald, D.D.; Urquidi-Macdonald, M.; Song, H.; Biaggio-Rocha, S.; Searson, P.

1991-11-01

This report summarizes the findings of our fundamental research program on passivity and passivity breakdown. During the past three and one half years in this program (including the three year incrementally-funded grant prior to the present grant), we developed and experimentally tested various physical models for the growth and breakdown of passive films on metal surfaces. These models belong to a general class termed ''point defects models'' (PDMs), in which the growth and breakdown of passive films are described in terms of the movement of anion and cation vacancies

Assessment of passive safety system of a Small Modular Reactor (SMR)

International Nuclear Information System (INIS)

Butt, Hassan Nawaz; Ilyas, Muhammad; Ahmad, Masroor; Aydogan, Fatih

2016-01-01

Highlights: • The MASLWR test facility has been modeled in RELAP5-SCDAP. The model is validated by comparing the simulation results with the experimental data. • Results obtained from various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core. • New scenarios are considered in which the effect of vent and sump recirculation valves failure has been investigated. • It is found from the results that continuous loss of inventory occurs due to lack of recirculation. • It is concluded that the high pressure vent valves in the MASLWR safety system require more redundancy. - Abstract: Innovative SMRs are designed with enhanced safety features based on lessons learnt from past experience of plant operation. Reliance on natural circulation and addition of passive safety systems made them inherently safe and simple in design. It is required to study reliability assessment of passive safety systems during postulated transients prior to their deployment on commercial scale. Test facilities and best estimate system codes are playing significant role in assessment of passive safety systems as well as in design, certification and evaluation of these innovative types of reactors. RELAP5 code is widely used for thermal-hydraulic analysis of nuclear reactors. In this work, the passive safety systems of Multi-Application Small Light Water (MASLWR) have been assessed. The complete loop of the MASLWR test facility has been modeled in RELAP5-SCDAP Mod 4.0. The RELAP5 model is validated by comparing the simulation results with the experimental data. Results obtained for various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core to avoid core heat up. Some of the components of passive safety system of MASLWR still rely on active power. Therefore, it was necessary to investigate their performance under failure

Design and analysis of EI core structured transverse flux linear reluctance actuator

OpenAIRE

FENERCİOĞLU, AHMET; AVŞAR, YUSUF

2015-01-01

In this study, an EI core linear actuator is proposed for horizontal movement systems. It is a transverse flux linear switched reluctance motor designed with an EI core structure geometrically. The actuator is configured into three phases and at a 6/4 pole ratio, and it has a stationary active stator along with a sliding passive translator. The stator consists of E cores and the translator consists of I cores. The actuator has a yokeless design because the stator and translator have no back i...

Experimental study of the passive flooding system in the WWER-1000 reactor

International Nuclear Information System (INIS)

Malyshev, A.B.; Efanov, A.D.; Kalyakin, S.G.

2002-01-01

The design solution of the passive flooding system in the WWER-1000 reactor core with the V-392 reactor facility and the scheme of the GE-2 large-scale thermohydraulic stand for substantiation of its functions are presented. The proposals, improving the efficiency of the system are developed on the basis of the experimental studies on the equipment input-output operational characteristics and the recommendations on the substantiation of the function of the reactor core flooding system are given [ru

Trends vs. reactor size of passive reactivity shutdown and control performance

International Nuclear Information System (INIS)

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

1988-01-01

The focus of the US advanced reactor program since the cancellation of CRBR has been on inherent safety and cost reduction. The notion is to so design the reactor that in the event of an off normal condition, it brings itself to a safe shutdown condition and removes decay heat by reliance on ''inherent processes'' i.e., without reliance on devices requiring switching and outside sources of power. Such a reactor design would offer the potential to eliminate costly ''Engineered Safety Features,'' to lower capital costs, and to assuage public unease concerning reactor safety. For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the passive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. 8 refs., 12 figs., 1 tab

Analysis of the AP600 core makeup tank experiments using the NOTRUMP code

International Nuclear Information System (INIS)

Cunningham, J.C.; Haberstroh, R.C.; Hochreiter, L.E.; Jaroszewicz, J.

1995-01-01

The AP600 design utilizes passive methods to perform core and containment cooling functions for a postulated loss of coolant. The core makeup tank (CMT) is an important feature of the AP600 passive safety system. The NOTRUMP code has been compared to the 300-series core makeup tank experiments. It has been observed that the code will capture the correct thermal-hydraulic behavior observed in the experiments. The correlations used for wall film condensation and convective heat transfer to the heated CMT liquid appear to be appropriate for these applications. The code will predict the rapid condensation and mixing thermal-hydraulic behavior observed in the 300-series tests. The NOTRUMP predictions can be noding-dependent since the condensation is extremely dependent on the amount of cold CMT liquid that mixes with the incoming steam flow

Reactor core cooling device

International Nuclear Information System (INIS)

Kobayashi, Masahiro.

1986-01-01

Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

International Nuclear Information System (INIS)

Yang, W.S.; Kim, T.K.; Grandy, C.

2007-01-01

This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

The planning of a passive seismic experiment: the Ketzin case

Science.gov (United States)

Rossi, G.; Petronio, L.

2009-04-01

In the last years, it has been recognized the importance of using microseismic activity data to gain information on the state and dynamics of a reservoir, notwithstanding the difficulties of recording, localizing the events, interpret them correctly, in terms of developing fractures, or thermal effects. The increasing number of CO2 storage experiments, with the necessity of providing efficient, economic, and long-term monitoring methods, both in the injection and post-injection phases, further encourage the development and improvement of recording and processing techniques. Microseismic signals are typically recorded with downhole sensors. Monitoring with surface sensors is problematic due to increased noise levels and signal attenuation particularly in the near surface. The actual detection distance depends on background noise conditions, seismic attenuation and the microseismic source strength. In the frame of the European project Co2ReMoVe and of the European Network of Excellence Co2GeoNet, a passive seismic experiment was planned in the Ketzin site for geological storage of CO2, a former gas store near Potsdam, object of the CO2SINK European project and inserted also in the European project Co2ReMoVe. Aim of the survey is to complement the CO2-SINK active seismic downhole experiments, adding precious information on the microseismicity induced by stress field changes at the reservoir level and in the overburden, due to the CO2 injection. The baseline survey was done in May 2008 by the Istituto Nazionale di Oceanografia e di Geofisica Sperimentale-OGS (Italy), with the support of the Deutsches GeoForschungsZentrum-GFZ (Germany) and the collaboration of the Institut für Geowissenschaftliche Gemeinschaftsaufgaben-GGA (Germany), shortly before the starting of the CO2 injection (June 30th 2008). A continuous monitoring (about 5 days) was performed by 2 downhole 3C geophones, and 3 surface 3C geophones located around the wells. This paper, based on the analysis of

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R+D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR 1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Meseth, J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R and D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

Grid-Connected Control Strategy of Five-level Inverter Based on Passive E-L Model

Directory of Open Access Journals (Sweden)

Tao Li

2017-10-01

Full Text Available At present, the research on five-level inverters mainly involves the modulation algorithm and topology, and few articles study the five-level inverter from the control strategy. In this paper, the nonlinear passivity-based control (PBC method is proposed for single-phase uninterruptible power supply inverters. The proposed PBC method is based on an energy shaping and damping injection idea, which is performed to regulate the energy flow of an inverter to a desired level and to assure global asymptotic stability, respectively. Furthermore, this paper presents a control algorithm based on the theory of passivity that gives an inverter in a photovoltaic system additional functions: power factor correction, harmonic currents compensation, and the ability to stabilize the system under varying injection damping. Finally, the effectiveness of the PBC method in terms of both stability and harmonic distortion is verified by the simulation and experiments under resistive and inductive loads.

Operating experience of natural circulation core cooling in boiling water reactors

International Nuclear Information System (INIS)

Kullberg, C.; Jones, K.; Heath, C.

1993-01-01

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

Reduced scale PWR passive safety system designing by genetic algorithms

International Nuclear Information System (INIS)

Cunha, Joao J. da; Alvim, Antonio Carlos M.; Lapa, Celso Marcelo Franklin

2007-01-01

This paper presents the concept of 'Design by Genetic Algorithms (DbyGA)', applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement

Survey of the passive safety systems of the BWR 1000 concept from SIEMENS

Energy Technology Data Exchange (ETDEWEB)

Mattern, J; Brettschuh, W; Palavecino, C [SIEMENS, Energieerzeugung, Offenbach (Germany)

1996-12-01

Through the use of passive safety systems and components for accident control in addition to the active systems required for plant operation, a higher degree of safety against core-endangering conditions is achieved which is no longer ruled by complex system engineering dependent on power supply and activation by I and C systems. A low core power density and large water inventories stored inside the reactor pressure vessel as well as inside and outside the containment ensure good plant behaviour in the event of transients or accidents. These passive safety systems - which required neither electric power to function nor I and C systems for actuation, being activated solely on the basis of changes in process variables such as water level, pressure and temperature - provide a grace period of more than 5 days after the onset of accident conditions before manual intervention becomes necessary. 8 figs.

Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

Energy Technology Data Exchange (ETDEWEB)

Jeon, In Seop, E-mail: jeoni@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Han, Sang Hoon, E-mail: shhan2@kaeri.re.kr [Advanced Research Group, Korea Atomic Energy Research Institute, 70 Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kang, Sang Hee, E-mail: sanghee.kang@khnp.co.kr [NSSS Design Group, Korea Hydro & Nuclear Power Co., Ltd., Central Research Institute, 70, 1312-beongil, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

2017-04-01

Highlights: • The novel F&B operation strategy with H-SIT and LPSI is developed. • The effectiveness of the H-SITs is verified using thermo-hydraulic simulations. • Success criteria considered for the new F&B operation strategy is identified. • A PSA model of APR+ reflecting the new F&B strategy with H-SIT is developed. • A risk analysis of the proposed F&B operation strategy is performed. - Abstract: While safety functions in current nuclear power plants are mainly provided by active safety systems, recently passive safety systems are being combined with the active systems to strengthen accident mitigation capability and therefore enhance overall plant safety. To this end, securing long-term cooling of the core is of particular importance. This study considers the hybrid safety injection tank (H-SIT), a passive injection system, as a target component to develop a long-term cooling strategy using active and passive systems concurrently. In the feed-and-bleed (F&B) operation, one of the important long-term cooling strategies to maintain core safety in pressurized water reactors, low pressure safety injection (LPSI) pumps are typically considered inoperable as depressurization is first required, which leads to core dry-out before reaching LPSI operable pressure. This study investigates whether H-SITs, with the important design feature of passive coolant injection under any pressure condition of the primary coolant system, can make up the core during depressurization thereby allowing LPSI pumps to be used in F&B operation as an additional means of long-term cooling. The effectiveness of the H-SITs is verified using thermal-hydraulic simulations, and based on the results a novel F&B operation strategy with H-SITs and LPSI pumps is developed. A probabilistic safety assessment (PSA) model is then developed in order to assess the risk effect of the suggested strategy. PSA results demonstrate that the proposed strategy lowers core damage frequency in the target

Passive decay heat removal by natural air convection after severe accidents

Energy Technology Data Exchange (ETDEWEB)

Erbacher, F.J.; Neitzel, H.J. [Forschungszentrum Karlsruhe Institut fur Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany); Cheng, X. [Technische Universitaet Karlsruhe Institut fur Stroemungslehre und Stroemungsmaschinen, Karlsruhe (Germany)

1995-09-01

The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

The effect of low capacity injection systems on transient initiated loss of vessel water injection at Browns Ferry unit one

International Nuclear Information System (INIS)

Ott, L.T.

1983-01-01

Probabilistic risk assessment (PRA) analyses have indicated the transient initiated loss of vessel water injection (TQUV sequence) to be a dominant accident scenario for BWR plants. The PRA studies assumed the low capacity injection systems to be unimportant in severe accidents. The results of a Severe Accident Sequence Analysis (SASA) Program study have shown that these systems are capable of preventing or significantly delaying core damage in a TQUV sequence

Passive solar technology

Energy Technology Data Exchange (ETDEWEB)

Watson, D

1981-04-01

The present status of passive solar technology is summarized, including passive solar heating, cooling and daylighting. The key roles of the passive solar system designer and of innovation in the building industry are described. After definitions of passive design and a summary of passive design principles are given, performance and costs of passive solar technology are discussed. Passive energy design concepts or methods are then considered in the context of the overall process by which building decisions are made to achieve the integration of new techniques into conventional design. (LEW).

Passive decay heat removal by sump cooling after core meltdown

International Nuclear Information System (INIS)

Knebel, J.U.; Mueller, U.

1996-01-01

This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives first measurement results of the 1:20 linearly scaled plane two-dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototype conditions

Performance Tests of Three Flow Distributors Using SMART-ITL with 1-Train CMT

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Shin, Yong-Cheol; Ko, Yung-Joo; Min, Kyoung-Ho; Ryu, Hyo Bong; Park, Jong-Kuk; Bang, Yun-Gon; Chae, Young-Jong; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Passive safety systems (PSSs) are key tools to remove the heat from the core or containment. Safety improvements for SMART have been studied since the Standard Design Approval (SDA) for SMART was certificated in 2012. Active safety systems such as safety injection pumps are replaced by a passive system, which is a kind of the gravity injection system with core makeup tanks (CMT) and safety injection tanks (SIT). All tanks for the passive safety systems are located higher than a pressurized reactor vessel, whose injection nozzles are located around the reactor coolant pumps (RCP). An Integral Test Loop for the SMART design (SMART-ITL) has been constructed and its commissioning tests finished in 2012. SMART-ITL is scaled down by the volume scaling methodology. Its height is conserved and its volume scale ratio is 1/49. The SMART-ITL has all fluid systems of SMART together with a break system and instruments. Recently, a test program to validate the performance of SMART Passive Safety System (PSS) was launched. A scaled-down test facility for SMART PSS was additionally installed at the existing SMART-ITL facility and a set of validation tests were performed. In this paper, the performance tests of the flow distributors using SMART-ITL with 1-train CMT will be discussed. A 1-train passive safety system including a CMT and SIT, which is operated only by gravity force, was additionally installed in the SMART-ITL to replace the active safety system for the SMART design. Several performance tests for the flow distributors were carried out to estimate a designed flow rate. 1. The peak flow rate in a hot test does not reach the value in a cold test, and the approaching time to peak is also delayed during the early stage of gravity injection. 2.. It is verified that the flow rate from a gravity injection depends on the differential pressure in the injection pipe line including a friction and form drag, which can be adjusted by controlling the resistance coefficient.

Modeling of reflood of severely damaged reactor core

International Nuclear Information System (INIS)

Bachrata, A.

2012-01-01

The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

Numerical simulation of boron injection in a BWR

Energy Technology Data Exchange (ETDEWEB)

Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

2010-02-15

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

Numerical simulation of boron injection in a BWR

International Nuclear Information System (INIS)

Tinoco, Hernan; Buchwald, Przemyslaw; Frid, Wiktor

2010-01-01

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several

Applied reliability assessment for the passive safety systems of nuclear power plants (NPPs) using system dynamics (SD)

International Nuclear Information System (INIS)

Kim, Yun Il; Woo, Tae Ho

2018-01-01

The passive system by the free-fall is investigated in the accident of nuclear power plants (NPPs). The complex algorithm of the system dynamics (SD) modeling is done in the passive cooling system. The nuclear passive system by free-fall is successfully modeled for the loss of coolant accident (LOCA). Conventional passive system of gravity or natural circulation is working only when the piping systems is in the good condition. The external coolant supply system is introduced in the case of the piping system failure. The water is poured into the reactor through the guiding piping or tube. If the explosion happens, the coolants could be showering into the reactor core and its building. New kind of passive system is expected successfully in the on-site black out where the drone could be operated by battery or engine.

Quantitation of passive cutaneous anaphylaxis (PCA) by using radiolabelled antigen

International Nuclear Information System (INIS)

Ring, J.; Seifert, J.; Brendel, W.

1978-01-01

The major problem of detecting reaginic antibody by passive cutaneous anaphylaxis (PCA) is the quantitation of the dye reaction. Radiolabelled antigen was used in an attempt to quantitate the PCA reaction (Radio-PCA). Antisera containing reaginic antibody against human serum albumin (HSA) were produced in rabbits. These antisera were injected into normal rabbit skin in different dilutions. Twentyfour hours later HSA was injected intravenously either with Evans Blue or as 125-I-HSA. Radioactivity found in antibody-containing skin was significantly higher than in control specimens containing saline or normal rabbit serum, as low as antiserum dilutions of 1:1,000. Compared with Evans Blue technique Radio-PCA was able to distinguish quantitatively between different antiserum dilutions at a higher level of statistical significance. (author)

Passive BWR integral LOCA testing at the Karlstein test facility INKA

Energy Technology Data Exchange (ETDEWEB)

Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

2014-05-15

KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

Modelling of liquid injection shutdown system (LISS) in ACR-1000

International Nuclear Information System (INIS)

Boubcher, M.; Colton, A.; Donnelly, J.V.

2008-01-01

Modelling of the Liquid Injection Shutdown System (LISS) in the ACR-1000 reactor core must account for the major phenomena that occur following its activation, namely the moderator hydraulics and core neutronics. The former requires modelling of the poison volumes, their time of entry into the reactor, and their propagation into the moderator after emission from the nozzle. The latter requires the reactivity worth of varying volumes and geometries of poisoned moderator fluid in order to simulate the reactivity effect of the injected poison. The time-dependent poison map is generated from hydraulic calculations, and then the neutronics data for standard geometries and concentrations is constructed using DRAGON. (author)

Passive shut-down of ITER plasma by Be evaporation

International Nuclear Information System (INIS)

Amano, Tsuneo.

1996-02-01

In an accident event where the cooling system of first wall of the ITER fails, the first wall temperature continues to rise as long as the ignited state of the core plasma persists. In this paper, a passive shut-down scheme of the ITER from this accident by evaporated Be from the first wall is examined. It is shown the estimated Be influx 5 10 24 /sec is sufficient to quench the ignition. (author)

A passive and active microwave-vector radiative transfer (PAM-VRT) model

International Nuclear Information System (INIS)

Yang, Jun; Min, Qilong

2015-01-01

A passive and active microwave vector radiative transfer (PAM-VRT) package has been developed. This fast and accurate forward microwave model, with flexible and versatile input and output components, self-consistently and realistically simulates measurements/radiation of passive and active microwave sensors. The core PAM-VRT, microwave radiative transfer model, consists of five modules: gas absorption (two line-by-line databases and four fast models); hydrometeor property of water droplets and ice (spherical and nonspherical) particles; surface emissivity (from Community Radiative Transfer Model (CRTM)); vector radiative transfer of successive order of scattering (VSOS); and passive and active microwave simulation. The PAM-VRT package has been validated against other existing models, demonstrating good accuracy. The PAM-VRT not only can be used to simulate or assimilate measurements of existing microwave sensors, but also can be used to simulate observation results at some new microwave sensors. - Highlights: • A novel microwave vector radiative transfer model is developed. • It can simulate passive and active microwave radiative transfer simultaneously. • It can be applied to simulate measurements for different types of viewing geometry. • The accuracy of this model has been validated against other existing models

CO2 Injectivity in Geological Storages: an Overview of Program and Results of the GeoCarbone-Injectivity Project

International Nuclear Information System (INIS)

Lombard, J.M.; Egermann, P.; Azaroual, M.; Pironon, J.; Broseta, D.; Egermann, P.; Munier, G.; Mouronval, G.

2010-01-01

The objective of the GeoCarbone-Injectivity project was to develop a methodology to study the complex phenomena involved in the near well bore region during CO 2 injection. This paper presents an overview of the program and results of the project, and some further necessary developments. The proposed methodology is based on experiments and simulations at the core scale, in order to understand (physical modelling and definition of constitutive laws) and quantify (calibration of simulation tools) the mechanisms involved in injectivity variations: fluid/rock interactions, transport mechanisms, geomechanical effects. These mechanisms and the associated parameters have then to be integrated in the models at the well bore scale. The methodology has been applied for the study of a potential injection of CO 2 in the Dogger geological formation of the Paris Basin, in collaboration with the other ANR GeoCarbone projects. (authors)

Evaluation report on SCTF Core-III tests S3-7 and S3-8

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi

1990-03-01

It has been said that the Emergency Core Cooling (ECC) water injected into the hot legs flows into the upper plenum and then falls back to the core (i.e. break-through) during reflood phase in a German type Pressurized Water Reactor (GPWR) with the combined-injection-type ECCS, and that the break-through occurs where the water temperature at the tie plate area is lower and subcooled. Based on this information two tests were conducted with the Slab Core Test Facility (SCTF) Core-III in order to investigate the effects of the water temperature distribution at the tie plate area on the break-through and the core cooling. In these tests, the subcooled ECC water was injected just above the Upper Core Support Plate (UCSP) in order to establish the desired water temperature distribution at the tie plate area. In one test (Test S3-7) the ECC water injection above the UCSP was performed above Bundles 3 and 4, and in the other test (Test S3-8) above Bundles 7 and 8 during initial 60 s a and then was changed to above Bundles 3 and 4. The test data were compared with those of Test S3-SH1, in which the injection was performed above Bundles 7 and 8 and the other test conditions were the same as in Tests S3-7 and S3-8. Analyzing these test data, the following has been found: The break-through occurs where the water temperature at the tie plate area is subcooled and the core cooling is enhanced significantly in the break-through region. The break-through location changes, with some time lag, following the change of the water temperature distribution at the tie plate area. Furthermore, the core cooling in the non-break-through regions is almost the same regardless of the location of the break-through. (author)

Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

International Nuclear Information System (INIS)

Murao, Y.; Iguchi, T.; Sugimoto, J.

1988-09-01

The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

Chaotic behavior in a system simulating the pressure balanced injection system. Analysis of passive safety reactor behavior. JAERI's nuclear research promotion program, H12-012 (Contract research)

Energy Technology Data Exchange (ETDEWEB)

Madarame, Haruki; Okamoto, Koji; Tanaka, Gentaro; Morimoto, Yuichiro [Tokyo Univ., School of Engineering, Tokyo (Japan); Sato, Akira [Yamagata Univ., Faculty of Engineering, Yonezawa, Yamagata (Japan); Kondou, Masaya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

The pressure Balanced Injection System (PBIS) was proposed in a passive safety reactor. Pressurizing Line (PL) connects the Reactor Vessel (RV) and the gas area in the Contain Vessel (CV), and Injected Line (IL) connects two vessels at relatively lower position. In an accident, the two lines are passively opened. The vapor generated by the residual heat pressed downward the water level in the RV. When the level is lower than the inlet of the PL, vapor is ejected into the CV through the PL attaining the pressure balance between the vessels. Then boron water in the CV is injected into the RV through the IL by the static head. This process is repeated by the succeeding vapor generation. In an experiment, the oscillating system was replaced by water column in a U-shaped duct. The vapor generation was simulated by cover gas supply to one end of the duct, while the other end was open to the atmosphere. When the water level reached a certain level, electromagnetic valves opened and the cover gas was ejected. The gas pressure decreased rapidly, resulting in a surface rise. When the water level reached another level, the valves closed. The cover gas pressure increased again, thus, gas ejection occurred intermittently. The interval of the gas ejection was not constant but fluctuated widely. Mere stochastic noise could hardly explain the large amplitude. Then was expressed the system using a set of linear equations. Various types of piecewise linear model were developed to examine the cause of the fluctuation. There appeared tangential bifurcation, period-doubling bifurcation, period-adding bifurcation and so on. The calculated interval exhibited chaotic features. Thus the cause of the fluctuation can be attributed to chaotic features of the system having switching. Since the piecewise linear model was highly simplified the behavior, a quantitative comparison between the calculation and the experiment was difficult. Therefore, numerical simulation code considering nonlinear

Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

International Nuclear Information System (INIS)

Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

2014-01-01

The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

Experiment on performance of upper head injection system with ROSA-II

International Nuclear Information System (INIS)

1978-05-01

Of the total 10 ROSA-II/UHI performance tests, 6 were reported previously. The rest are presented and discussion is made on the effects of heat generation in the core and UHI injection and repeatability of experiments. In addition, the following are described: (1) Pressure spikes observed in the upper head after sudden stoppage of UHI injection, and (2) discharge flow oscillation possibly due to UHI water injection into the upper plenum. (auth.)

Fuelling effect of tangential compact toroid injection in STOR-M Tokamak

Energy Technology Data Exchange (ETDEWEB)

Onchi, T.; Liu, Y., E-mail: tao668@mail.usask.ca [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Dreval, M. [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Inst. of Plasma Physics NSC KIPT, Kharkov (Ukraine); McColl, D. [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Asai, T. [Inst. of Plasma Physics NSC KIPT, Kharkov (Ukraine); Wolfe, S. [Nihon Univ., Dept. of Physics, Tokyo (Japan); Xiao, C.; Hirose, A. [Univ. of Saskatchewan, Saskatoon, Saskatchewan (Canada)

2012-07-01

Compact torus injection (CTI) is the only known candidate for directly fuelling the core of a tokamak fusion reactor. Compact torus (CT) injection into the STOR-M tokamak has induced improved confinement accompanied by an increase in the electron density, reduction in Hα emission, and suppression of the saw-tooth oscillations. The measured change in the toroidal flow velocity following tangential CTI has demonstrated momentum injection into the STOR-M plasma. (author)

Inhibition of PKMzeta in nucleus accumbens core abolishes long-term drug reward memory.

Science.gov (United States)

Li, Yan-qin; Xue, Yan-xue; He, Ying-ying; Li, Fang-qiong; Xue, Li-fen; Xu, Chun-mei; Sacktor, Todd Charlton; Shaham, Yavin; Lu, Lin

2011-04-06

During abstinence, memories of drug-associated cues persist for many months, and exposure to these cues often provokes relapse to drug use. The mechanisms underlying the maintenance of these memories are unknown. A constitutively active atypical protein kinase C (PKC) isozyme, protein kinase M ζ (PKMζ), is required for maintenance of spatial memory, conditioned taste aversion, and other memory forms. We used conditioned place preference (CPP) and conditioned place aversion (CPA) procedures to study the role of nucleus accumbens PKMζ in the maintenance of drug reward and aversion memories in rats. Morphine CPP training (10 mg/kg, 4 pairings) increased PKMζ levels in accumbens core but not shell. Injections of the PKMζ inhibitor ζ inhibitory peptide (ZIP) into accumbens core but not shell after CPP training blocked morphine CPP expression for up to 14 d after injections. This effect was mimicked by the PKC inhibitor chelerythrine, which inhibits PKMζ, but not by the conventional and novel PKC inhibitor staurosporine, which does not effectively inhibit PKMζ. ZIP injections into accumbens core after training also blocked the expression of cocaine (10 mg/kg) and high-fat food CPP but had no effect on CPA induced by naloxone-precipitated morphine withdrawal. Accumbens core injections of Tat-GluR2(3Y), which inhibits GluR2-dependent AMPA receptor endocytosis, prevented the impairment in morphine CPP induced by local ZIP injections, indicating that the persistent effect of PKMζ is on GluR2-containing AMPA receptors. Results indicate that PKMζ activity in accumbens core is a critical cellular substrate for the maintenance of memories of relapse-provoking reward cues during prolonged abstinence periods.

Core catcher concepts future PWR-Plants

International Nuclear Information System (INIS)

Alsmeyer, H.; Werle, H.

1994-01-01

Light water reactors of the next generation should have still greater passive safety, even in the most serious accidents. This includes the long term safe inclusion of the core inventory in the case of core meltdown accidents. The three concepts for cooling the liquefied core outside the reactor pressure vessel examined by KfK should remove the post-shutdown heat by direct contact of the melt with water. The geometric distribution of the melt increases its surface area, so that favourable conditions for heat removal from the poorly thermally-conducting melt are created and complete quick solidification occurs. The experiments examine both the relocation and distribution mechanisms of the melt and the reactions occurring when water enters. As strong interaction is possible on direct contact of the melt with water, an important aim is experimental determination and limitation of any resulting mechanical stresses. (orig./HP) [de

Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

International Nuclear Information System (INIS)

Araya, F.

1995-01-01

After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

An update on passive correctors for the SSC dipole magnets

International Nuclear Information System (INIS)

Green, M.A.

1991-05-01

The concept of correction of the magnetization sextupole became a topic of discussion as soon as it was realized that superconductor magnetization could have a serious effect on the SSC beam during injection. Several methods of correction were proposed. These included (1) correction with active bore tube windings like those on the HERA machine which correct out magnetization sextupole and the sextupole due to iron saturation, (2) correction with persistent sextupole windings mounted on the bore tube (3) correction using passive superconductor (4) correction using ferromagnetic material, and (5) correction using oriented magnetized materials. This report deals with the use of passive superconductor to correct the magnetization sextupole. Two basic methods are explored in this report: (1) One can correct the magnetization sextupole by changing the diameter of the superconductor filaments in one or more blocks of the SSC dipole. (2) One can correct the magnetization sextupole and decapole by mounting passive superconducting wires on the inside of the SSC dipole coil bore. In addition, an assessment of the contribution of each conductor in the dipole to the magnetization sextupole and decapole is shown. 38 refs, 25 figs., 15 tabs

Assessment of core thermo-hydrodynamic models of REFLA-1D with CCTF data

International Nuclear Information System (INIS)

Okubo, Tsutomu; Murao, Yoshio

1983-07-01

In order to assess the core thermo-hydrodynamic models of REFLA-1D/MODE3, which is the latest version of REFLA-1D, several calculations of the core thermo-hydrodynamics have been performed for the CCTF Core-I series tests. The measured initial and boundary conditions were used for these calculations. The calculational results showed that the water accumulation model of Case 2 could predict the CCTF results fairly well as it could for the JAERI small scale facility. The calculated results for the base case and the EM tests were in good agreement with the CCTF data. The parameter effects, such as system pressure, initial clad temperature, Acc injection rate, LPCI injection rate and initial down-comer wall temperature, were predicted correctly, except for the high system pressure and the high LPCI injection rate tests. (author)

Self-passivation Rule and the Effect of Post-treatment in GBs of Solar Cell Materials

Science.gov (United States)

Liu, Chengyan; Chen, Shiyou; Xiang, Hongjun; Gong, Xingao

Grain boundaries (GBs) existing in polycrystalline semiconductors alloys inducing a great deal of deep defect levels are usually harmful to cells' photovoltaic performance. Experimental and theoretical investigations verified that these defect levels come from the GBs' dangling bonds. We find that, the defect levels in anion core of GB can be passivated by its cations, called by self-passivation. For instance, the post-treated by CdCl2, Cd can eliminate the defect levels by saturating Te dangling bonds in the grain boundary of CdTe. We verify that the idea of self-passivation rule can perfectly explain the benign GBs of CISe and CZTS by sodium treatment. The present work reveals a general mechanism about how dopants in GBs eliminate the defect states through passivating the dangling bonds in covalent polycrystalline semiconductors, and sheds light on how to passivate dangling bonds in GBs with alterative processes. National Science Foundation of China, international collaboration project of MOST, Pujiang plan, Program for Professor of Special Appointment (Eastern Scholar), and Shanghai Rising-star program.

Experimental Investigation In The PANDA Facility Of The Impact Of A Hydrogen Release On Passive Containment Cooling Systems

Energy Technology Data Exchange (ETDEWEB)

Auban, O.; Paladino, D.; Candreia, P.; Huggenberger, M.; Strassberger, H.J

2003-03-01

The large-scale, thermal-hydraulics PANDA facility has been used in the past for investigating passive decay-heat removal systems and related containment phenomena for a number of next-generation Light Water Reactors. Passive Containment Condenser (PCC) systems operate by transferring heat via steam condensation from inside the containment to outside, and serve to mitigate pressure build-up in the event of steam discharge from the primary circuit. Five new integral tests have recently been performed in the context of the 5th European Framework Program project TEMPEST. One main objective was to assess the influence of a light gas (hydrogen) on the performance of the Passive Containment Cooling System (PCCS). Hydrogen release in the case of a severe accident is simulated in PANDA by injecting helium with steam into the Drywell. In addition, the impact of a new accident-mitigating design feature, the Drywell Gas Re-Circulation System (DGRS), on the long-term containment behaviour was tested. Another important objective was also to provide relevant data for the validation of modern 3/4 mainly Computational Fluid Dynamics (CFD) 3/4 codes. The paper reports main observations from two of these new integral tests in which standard PANDA instrumentation has provided important data concerning system response when helium is released in the course of the transient. The results show that some important stratification phenomena have occurred, as a result of the buoyant flow generated by the helium injection. The resulting temperature and concentration measurements show how helium is being distributed in the Drywell (DW) volume during the helium injection phase, and how helium is retained or vented out of these vessels once the injection stops. The paper includes some discussion concerning the influence of gas mixing and stratification and the effect of the DGRS on PCC performance and system pressure build-up. (author)

Analysis and prevention of water hammer for the emergency core cooling system

International Nuclear Information System (INIS)

Zhao Jun

2008-01-01

Emergency core cooling system (ECCS) is an engineered safety feature of nuclear power plant. If the water hammer happens during ECCS injection, the piping system may be broken. It will cause loss of ECC system and affect the safety of reactor core. Based on the functions and characteristics of ECCS and the theory of water hammer, the paper analyzed the potential risk of water hammer in ECCS in Qinshan III, and proposed modifications to prevent the water-hammer damage during ECCS injection. (authors)

Passive body heating improves sleep patterns in female patients with fibromyalgia

Directory of Open Access Journals (Sweden)

Andressa Silva

2013-01-01

Full Text Available OBJECTIVE: To assess the effect of passive body heating on the sleep patterns of patients with fibromyalgia. METHODS: Six menopausal women diagnosed with fibromyalgia according to the criteria determined by the American College of Rheumatology were included. All women underwent passive immersion in a warm bath at a temperature of 36 ±1 °C for 15 sessions of 30 minutes each over a period of three weeks. Their sleep patterns were assessed by polysomnography at the following time-points: pre-intervention (baseline, the first day of the intervention (acute, the last day of the intervention (chronic, and three weeks after the end of the intervention (follow-up. Core body temperature was evaluated by a thermistor pill during the baseline, acute, chronic, and follow-up periods. The impact of this treatment on fibromyalgia was assessed via a specific questionnaire termed the Fibromyalgia Impact Questionnaire. RESULTS: Sleep latency, rapid eye movement sleep latency and slow wave sleep were significantly reduced in the chronic and acute conditions compared with baseline. Sleep efficiency was significantly increased during the chronic condition, and the awakening index was reduced at the chronic and follow-up time points relative to the baseline values. No significant differences were observed in total sleep time, time in sleep stages 1 or 2 or rapid eye movement sleep percentage. The core body temperature and Fibromyalgia Impact Questionnaire responses did not significantly change over the course of the study. CONCLUSION: Passive body heating had a positive effect on the sleep patterns of women with fibromyalgia.

Passive body heating improves sleep patterns in female patients with fibromyalgia

Science.gov (United States)

Silva, Andressa; de Queiroz, Sandra Souza; Andersen, Monica Levy; Mônico-Neto, Marcos; da Silveira Campos, Raquel Munhoz; Roizenblatt, Suely; Tufik, Sergio; de Mello, Marco Túlio

2013-01-01

OBJECTIVE: To assess the effect of passive body heating on the sleep patterns of patients with fibromyalgia. METHODS: Six menopausal women diagnosed with fibromyalgia according to the criteria determined by the American College of Rheumatology were included. All women underwent passive immersion in a warm bath at a temperature of 36±1°C for 15 sessions of 30 minutes each over a period of three weeks. Their sleep patterns were assessed by polysomnography at the following time-points: pre-intervention (baseline), the first day of the intervention (acute), the last day of the intervention (chronic), and three weeks after the end of the intervention (follow-up). Core body temperature was evaluated by a thermistor pill during the baseline, acute, chronic, and follow-up periods. The impact of this treatment on fibromyalgia was assessed via a specific questionnaire termed the Fibromyalgia Impact Questionnaire. RESULTS: Sleep latency, rapid eye movement sleep latency and slow wave sleep were significantly reduced in the chronic and acute conditions compared with baseline. Sleep efficiency was significantly increased during the chronic condition, and the awakening index was reduced at the chronic and follow-up time points relative to the baseline values. No significant differences were observed in total sleep time, time in sleep stages 1 or 2 or rapid eye movement sleep percentage. The core body temperature and Fibromyalgia Impact Questionnaire responses did not significantly change over the course of the study. CONCLUSION: Passive body heating had a positive effect on the sleep patterns of women with fibromyalgia. PMID:23525306

Design and analysis of a new passive residual heat removal system

Energy Technology Data Exchange (ETDEWEB)

Lv, Xing [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Peng, Minjun, E-mail: heupmj@163.com [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Yuan, Xiao [Guangxi Fangchenggang Nuclear Power Co., Ltd (China); Xia, Genglei [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China)

2016-07-15

Highlights: • An air cooling passive residual heat removal System (PRHRs) is designed. • Using RELAP5/MOD3.4 code to analyze the operation characteristics of the PRHRs. • Noncondensable gas is used to simulate the hydrodynamic behavior in the air cooling tower. • The natural circulations could respectively establish in the primary circuit and the PRHRs circuit. • The PRHRs could remove the residual heat effectively. - Abstract: The inherent safety functions will mitigate the consequences of the accidents, and it can be accomplished through the passive safety systems which employed in the typical pressurized water reactor (PWR). In this paper, a new passive residual heat removal system (PRHRS) is designed for a typical nuclear power plant. PRHRS consists of a steam generator (SG), a cooling tank with two groups of cooling pipes, an air-cooling heat exchanger (AHX), an air-cooling tower, corresponding pipes and valves. The cooling tank which works as an intermediate buffer device is used to transfer the core decay heat to the AHX, and then the core decay heat will be removed to the atmosphere finally. The RELAP5/MOD3.4 code is used to analyze the operation characteristics of PRHRS and the primary loop system. It shows PRHRS could remove the decay heat from the primary loop effectively, and the natural circulations can be established in the primary circuit and the PRHRS circuit respectively. Furthermore, the sensitivity study has also been done to research the effect of various factors on the heat removal capacity.

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)

Compositional properties of passivity

NARCIS (Netherlands)

Kerber, Florian; van der Schaft, Arjan

2011-01-01

The classical passivity theorem states that the negative feedback interconnection of passive systems is again passive. The converse statement, - passivity of the interconnected system implies passivity of the subsystems -, turns out to be equally valid. This result implies that among all feasible

Pellet injection and plasma behavior simulation code PEPSI

International Nuclear Information System (INIS)

Takase, Haruhiko; Tobita, Kenji; Nishio, Satoshi

2003-08-01

Fueling is one of the major issues on design of nuclear fusion reactor and the injection of solid hydrogen pellet to the core plasma is a useful method. On the design of a nuclear fusion reactor, it is necessary to determine requirements on the pellet size, the number of pellets, the injection speed and the injection cycle. PEllet injection and Plasma behavior SImulation code PEPSI has been developed to assess these parameters. PEPSI has two special features: 1) Adopting two numerical pellet models, Parks model and Strauss model, 2) Calculating fusion power and other plasma parameters in combination with a time-dependent one-dimensional transport model. This report describes the numerical models, numerical scheme, sequence of calculation, list of subroutines, list of variables and an example of calculation. (author)

Electrical properties of spherical dipole antennas with lossy material cores

DEFF Research Database (Denmark)

Hansen, Troels Vejle; Kim, Oleksiy S.; Breinbjerg, Olav

2012-01-01

A spherical magnetic dipole antenna with a linear, isotropic, homogenous, passive, and lossy material core is modeled analytically, and closed form expressions are given for the internally stored magnetic and electric energies, the radiation efficiency, and radiation quality factor. This model...... and all the provided expressions are exact and valid for arbitrary core sizes, permeability, permittivity, electric and magnetic loss tangents. Arbitrary dispersion models for both permeability and permittivity can be applied. In addition, we present an investigation for an antenna of fixed electrical...

Acute lethal toxicity following passive immunization for treatment of murine cryptococcosis.

Science.gov (United States)

Savoy, A C; Lupan, D M; Manalo, P B; Roberts, J S; Schlageter, A M; Weinhold, L C; Kozel, T R

1997-01-01

Passive immunization with monoclonal antibodies (MAbs) specific for the major capsular polysaccharide of Cryptococcus neoformans alters the course of murine cryptococcosis. During studies of passive immunization for treatment of murine cryptococcosis, we noted the occurrence of an acute, lethal toxicity. Toxicity was characterized by scratching, lethargy, respiratory distress, collapse, and death within 20 to 60 min after injection of antibody. The toxic effect was observed only in mice with a cryptococcal infection and was reduced or absent in the early and late stages of disease. The clinical course and histopathology were consistent with those for shock. There was considerable variation between mouse strains in susceptibility to toxicity. Swiss Webster mice from the Charles River colony were most susceptible, followed by C3H/He, BALB/c, and C57BL/6 mice. DBA/2 mice and Swiss Webster mice from the Simonsen colony were resistant. Acute toxicity was mimicked by injection of preformed complexes of MAb and purified polysaccharide. The toxic effect was also produced by injection of MAbs into mice that were preloaded with polysaccharide. The toxic effect was not blocked by treatment of mice with chloropheniramine or anti-tumor necrosis factor alpha antibodies or by depletion of complement components via pretreatment with cobra venom factor. Toxicity was reduced by treatment of mice with high doses of epinephrine, dexamethasone, or chlorpromazine. Finally, the toxic effect was completely blocked by treatment of mice with the platelet-activating factor antagonist WEB 2170 BS or by pretreatment of mice with the liposome-encapsulated drug dichloromethylene diphosphonate, a procedure which depletes macrophages from the spleen and liver. PMID:9125564

Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

Science.gov (United States)

Peterson, Per F.

2013-05-14

A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

Energy Technology Data Exchange (ETDEWEB)

Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

2016-03-15

Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

International Nuclear Information System (INIS)

Park, Rae-Joon; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

2016-01-01

Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m"2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

Minimally invasive injectable short nanofibers of poly(glycerol sebacate) for cardiac tissue engineering

International Nuclear Information System (INIS)

Ravichandran, Rajeswari; Venugopal, Jayarama Reddy; Sundarrajan, Subramanian; Mukherjee, Shayanti; Sridhar, Radhakrishnan; Ramakrishna, Seeram

2012-01-01

Myocardial tissue lacks the ability to appreciably regenerate itself following myocardial infarction (MI) which ultimately results in heart failure. Current therapies can only retard the progression of disease and hence tissue engineering strategies are required to facilitate the engineering of a suitable biomaterial to repair MI. The aim of this study was to investigate the in vitro properties of an injectable biomaterial for the regeneration of infarcted myocardium. Fabrication of core/shell fibers was by co-axial electrospinning, with poly(glycerol sebacate) (PGS) as core material and poly-l-lactic acid (PLLA) as shell material. The PLLA was removed by treatment of the PGS/PLLA core/shell fibers with DCM:hexane (2:1) to obtain PGS short fibers. These PGS short fibers offer the advantage of providing a minimally invasive injectable technique for the regeneration of infarcted myocardium. The scaffolds were characterized by SEM, FTIR and contact angle and cell–scaffold interactions using cardiomyocytes. The results showed that the cardiac marker proteins actinin, troponin, myosin heavy chain and connexin 43 were expressed more on short PGS fibers compared to PLLA nanofibers. We hypothesized that the injection of cells along with short PGS fibers would increase cell transplant retention and survival within the infarct, compared to the standard cell injection system. (paper)

Passive fault current limiting device

Science.gov (United States)

Evans, Daniel J.; Cha, Yung S.

1999-01-01

A passive current limiting device and isolator is particularly adapted for use at high power levels for limiting excessive currents in a circuit in a fault condition such as an electrical short. The current limiting device comprises a magnetic core wound with two magnetically opposed, parallel connected coils of copper, a high temperature superconductor or other electrically conducting material, and a fault element connected in series with one of the coils. Under normal operating conditions, the magnetic flux density produced by the two coils cancel each other. Under a fault condition, the fault element is triggered to cause an imbalance in the magnetic flux density between the two coils which results in an increase in the impedance in the coils. While the fault element may be a separate current limiter, switch, fuse, bimetal strip or the like, it preferably is a superconductor current limiter conducting one-half of the current load compared to the same limiter wired to carry the total current of the circuit. The major voltage during a fault condition is in the coils wound on the common core in a preferred embodiment.

Passive immunization of Pacific herring against viral hemorrhagic septicemia.

Science.gov (United States)

Hershberger, P.K.; Gregg, J.L.; Grady, C.A.; LaPatra, S.E.; Winton, J.R.

2011-01-01

The plasma of Pacific herring Clupea pallasii that survived laboratory-induced viral hemorrhagic septicemia (VHS) epizootics contained humoral substances that, when injected into naive animals, conferred passive immunity against the disease. Among groups exposed to viral hemorrhagic septicemia virus (VHSV), injection of donor plasma from VHS survivors resulted in significantly greater survival (50%) and significantly lower tissue titers (1.5 x 10(5) plaque-forming units [PFU]/g) than the injection of plasma from VHSV-naive donors (6% survival; 3.7 x 10(6) PFU/g). Additionally, the magnitude of the protective immune response increased during the postexposure period; plasma that was collected from survivors at 123 d postexposure (931 degree-days) provided greater protection than plasma collected from survivors at 60 d postexposure (409 degree-days). These results provide proof of concept that the VHSV exposure history of Pacific herring populations can be determined post hoc; furthermore, the results can be used as the foundation for developing additional high-throughput diagnostic techniques that may be effective at quantifying herd immunity and forecasting the potential for future VHS epizootics in populations of wild Pacific herring.

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

1999-01-01

Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

Optical power equalization for upstream traffic with injection-locked Fabry-Perot lasers in TDM-PON

Science.gov (United States)

Huang, Ting-Tsan; Sheu, Lih-Gen; Chi, Sien

2010-10-01

An optical power equalization of upstream traffic in time-division-multiplexed passive optical network (TDM-PON) based on injection-locked Fabry-Perot lasers has been experimentally investigated. The upstream transmitters with stable spectrum are achieved by using an external injection light source in the optical line terminal (OLT). The different upstream powers can be equalized by injection locking a Fabry-Perot laser diode (FP-LD) biased below threshold current in OLT. The dynamic upstream power range from - 8.5 to - 19.5 db m is reduced to a 1.6 dB maximal power variation, when the uplink signal is directly modulated at 1.25 Gb/s.

Station blackout core damage frequency in an advanced nuclear reactor

International Nuclear Information System (INIS)

Carvalho, Luiz Sergio de

2004-01-01

Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

International Nuclear Information System (INIS)

Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

2003-01-01

emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments

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.

Integration of the functional reliability of two passive safety systems to mitigate a SBLOCA+BO in a CAREM-like reactor PSA

Energy Technology Data Exchange (ETDEWEB)

Mezio, Federico, E-mail: federico.mezio@cab.cnea.gov.ar [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Grinberg, Mariela [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina); Lorenzo, Gabriel [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Giménez, Marcelo [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina)

2014-04-01

Highlights: • An estimation of the Functional Unreliability was performed using RMPS methodology. • The methodology uses an improved response surface in order to estimate the FU. • The FU may become relevant to be analyzed in the Passive Safety Systems. • There were proposed two ways to incorporate the FU into an APS. - Abstract: This paper describes a case study of a methodological approach for assessing the functional reliability of passive safety systems (PSS) and its treatment within a probabilistic safety assessment (PSA). The functional unreliability (FU) can be understood as the failure probability of PSS to fulfill its mission due to the impairment of the related passive safety function. The safety function accomplishment is characterized and quantified by a performance indicator (PI), which is a measure of how far the system is from verifying its mission. PI uncertainties are estimated from uncertainty propagation of selected parameters. A methodology based on the reliability methodology for passive system (RMPS) one is used to estimate the FU associated to the isolation condensers (ICs) in combination with the accumulators (medium pressure injection system) of a CAREM-like integral advanced reactor. A small break loss of coolant accident with black-out is selected as an evaluation case. This implies success of reactor shut-down (inherent) and failure of residual heat removal by active systems. The safety function to accomplish is to refill the reactor pressure vessel (RPV) in order to avoid core damage. For this case, to allow the discharge of accumulators into RPV, the pressure must be reduced by the IC. The methodology for passive safety function assessment considers uncertainties in code parameters, besides uncertainties in engineering parameters (design, construction, operation and maintenance), in order to perform Monte Carlo simulations based on best estimate (B-E) plant model. Then, response surfaces based on PI are used for improving the

Cardiac lymphoscintigraphy following closed-chest catheter injection of radiolabeled colloid into the myocardium of dogs: concise communication

International Nuclear Information System (INIS)

Osbakken, M.D.; Kopiwoda, S.Y.; Swan, A.; Castronovo, F.P.; Strauss, H.W.

1982-01-01

A catheter technique for injection of radiolabeled colloids into the myocardium was developed and tested in a series of 15 dogs. A multipurpose angiographic catheter was modified to permit an inner core of PE-50 polyethylene tubing, tipped with a 23-gage needle, to pass through the lumen for intra-myocardial injection of radiocolloids. For injection of the left ventricle, the catheter is introduced through the femoral artery: for the right ventricle, the femoral vein. The catheter advanced under fluoroscopy until the desired surface for injection is reached. The inner core is then extended to lodge the needle in the endocardium. A mixture of Renografin (to confirm the endocardial injection site) and radiolabeled colloid was injected in 13 animals. Ten minutes after injection, scintigraphy was begun and continued for up to 6 hr. In three dogs the procedure was repeated 3 or 4 times. From two to five nodes were visible in all animals, irrespective of whether the right or left ventricular myocardium was injected. In two animals the injection was given intravenously, and no nodes were seen. These data suggest that cardiac lymphatic drainage can be studied with a catheter injection method

Cardiac lymphoscintigraphy following closed-chest catheter injection of radiolabeled colloid into the myocardium of dogs: concise communication

International Nuclear Information System (INIS)

Osbakken, M.D.; Kopiwoda, S.Y.; Swan, A.; Castronovo, F.P.; Strauss, H.W.

1982-01-01

A catheter technique for injection of radiolabeled colloids into the myocardium was developed and tested in a series of 15 dogs. A multipurpose angiographic catheter was modified to permit an inner core of PE-50 polyethylene tubing, tipped with a 23-gage needle, to pass through the lumen for intra-myocardial injection of the femoral artery: for the right ventricle, the femoral vein. The catheter advanced under fluoroscopy until the desired surface for injection is reached. The inner core is then extended to lodge the needle in the endocardium. A mixture of Renogratin (to confirm the endocardial injection site) and radiolabeled colloid was injected in up to 6 hr. In three dogs the procedure was repeated 3 or 4 times. From two to five nodes were visible in all animals, irrespective of whether the right or left ventricular myocardium was injected. In two animals the injection was given intravenously, and no nodes were seen. These data suggest that cardiac lymphatic drainage can be studied with a catheter injection method

Thick-Film and LTCC Passive Components for High-Temperature Electronics

Directory of Open Access Journals (Sweden)

A. Dziedzic

2013-04-01

Full Text Available At this very moment an increasing interest in the field of high-temperature electronics is observed. This is a result of development in the area of wide-band semiconductors’ engineering but this also generates needs for passives with appropriate characteristics. This paper presents fabrication as well as electrical and stability properties of passive components (resistors, capacitors, inductors made in thick-film or Low-Temperature Co-fired Ceramics (LTCC technologies fulfilling demands of high-temperature electronics. Passives with standard dimensions usually are prepared by screen-printing whereas combination of standard screen-printing with photolithography or laser shaping are recommenced for fabrication of micropassives. Attainment of proper characteristics versus temperature as well as satisfactory long-term high-temperature stability of micropassives is more difficult than for structures with typical dimensions for thick-film and LTCC technologies because of increase of interfacial processes’ importance. However it is shown that proper selection of thick-film inks together with proper deposition method permit to prepare thick-film micropassives (microresistors, air-cored microinductors and interdigital microcapacitors suitable for the temperature range between 150°C and 400°C.

Temperature Dependence and Magnetic Properties of Injection Molding Tool Materials Used in Induction Heating

DEFF Research Database (Denmark)

Guerrier, Patrick; Nielsen, Kaspar Kirstein; Hattel, Jesper Henri

2015-01-01

To analyze the heating phase of an induction heated injection molding tool precisely, the temperature-dependent magnetic properties, B–H curves, and the hysteresis loss are necessary for the molding tool materials. Hence, injection molding tool steels, core materials among other materials have...

Passive decay heat removal by natural circulation

International Nuclear Information System (INIS)

Vijayan, P.K.; Venkat Raj, V.; Kakodkar, A.; Mehta, S.K.

1990-01-01

The standardised 235 MWe PHWRs being built in India are the pressure tube type, heavy water moderated, heavy water cooled and natural uranium fuelled reactors. Several passive safety features are incorporated in these reactors. These include: (1) Containment pressure reduction and fission product trapping with the help of suppression pool following LOCA. (2) Emergency coolant injection by means of accumulators. (3) Large heat sink provided by the low temperature moderator under accident conditions. (4) Low excess reactivity, through the use of natural uranium fuel and on power fuelling. (5) Residual heat removal by means of natural circulation, etc. of which the last item is the subject matter of this report. (author). 8 refs, 10 figs

Ideal magnetohydrodynamic simulations of unmagnetized dense plasma jet injection into a hot strongly magnetized plasma

OpenAIRE

Liu, Wei; Hsu, Scott C.

2010-01-01

We present results from three-dimensional ideal magnetohydrodynamic simulations of unmagnetized dense plasma jet injection into a uniform hot strongly magnetized plasma, with the aim of providing insight into core fueling of a tokamak with parameters relevant for ITER and NSTX (National Spherical Torus Experiment). Unmagnetized dense plasma jet injection is similar to compact toroid injection but with much higher plasma density and total mass, and consequently lower required injection velocit...

Microcellular injection molding process for producing lightweight thermoplastic polyurethane with customizable properties

Science.gov (United States)

Ellingham, Thomas; Kharbas, Hrishikesh; Manitiu, Mihai; Scholz, Guenter; Turng, Lih-Sheng

2018-03-01

A three-stage molding process involving microcellular injection molding with core retraction and an "out-of-mold" expansion was developed to manufacture thermoplastic polyurethane into lightweight foams of varying local densities, microstructures, and mechanical properties in the same microcellular injection molded part. Two stages of cavity expansion through sequential core retractions and a third expansion in a separate mold at an elevated temperature were carried out. The densities varied from 0.25 to 0.42 g/cm3 (77% to 62% weight reduction). The mechanical properties varied as well. Cyclic compressive strengths and hysteresis loss ratios, together with the microstructures, were characterized and reported.

Interband optical pulse injection locking of quantum dot mode-locked semiconductor laser.

Science.gov (United States)

Kim, Jimyung; Delfyett, Peter J

2008-07-21

We experimentally demonstrate optical clock recovery from quantum dot mode-locked semiconductor lasers by interband optical pulse injection locking. The passively mode-locked slave laser oscillating on the ground state or the first excited state transition is locked through the injection of optical pulses generated via the opposite transition bands, i.e. the first excited state or the ground state transition from the hybridly mode-locked master laser, respectively. When an optical pulse train generated via the first excited state from the master laser is injected to the slave laser oscillating via ground state, the slave laser shows an asymmetric locking bandwidth around the nominal repetition rate of the slave laser. In the reverse injection case of, i.e. the ground state (master laser) to the first excited state (slave laser), the slave laser does not lock even though both lasers oscillate at the same cavity frequency. In this case, the slave laser only locks to higher injection rates as compared to its own nominal repetition rate, and also shows a large locking bandwidth of 6.7 MHz.

The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

International Nuclear Information System (INIS)

Hill, D.J.; Ragland, W.A.; Roglans, J.

1998-01-01

This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

Energy Technology Data Exchange (ETDEWEB)

Hill, D J; Ragland, W A; Roglans, J

1998-11-01

This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10{sup -6} yr{sup -1}, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10{sup -6} yr{sup -1} and the contribution of seismic events is 1.7 10{sup -5} yr{sup -1}. Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

The EBR-II probabilistic risk assessment lessons learned regarding passive safety

International Nuclear Information System (INIS)

Hill, D.J.; Ragland, W.A.; Roglans, J.

1994-01-01

This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

Thermal stability studies on atomically clean and sulphur passivated InGaAs surfaces

Energy Technology Data Exchange (ETDEWEB)

Chauhan, Lalit; Hughes, Greg [School of Physical Sciences, Dublin City University, Glasnevin, Dublin 9 (Ireland)

2013-03-15

High resolution synchrotron radiation core level photoemission measurements have been used to study the high temperature stability of sulphur passivated InGaAs surfaces and comparisons made with atomically clean surfaces subjected to the same annealing temperatures. Sulphur passivation of clean InGaAs surfaces prepared by the thermal removal of an arsenic capping layer was carried out using an in situ molecular sulphur treatment in ultra high vacuum. The elemental composition of the surfaces of these materials was measured at a series of annealing temperatures up to 530 C. Following a 480 C anneal In:Ga ratio was found to have dropped by 33% on sulphur passivated surface indicating a significant loss of indium, while no drop in indium signal was recorded at this temperature on the atomically InGaAs surface. No significant change in the As surface concentration was measured at this temperature. These results reflect the reduced thermal stability of the sulphur passivated InGaAs compared to the atomically clean surface which has implications for device fabrication. (Copyright copyright 2013 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

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.

Scoping analyses for the safety injection system configuration for Korean next generation reactor

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Song, Jin Ho; Park, Jong Kyoon

1996-01-01

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are performed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSl pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SlT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

2016-01-01

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

Injection system of compact SR light source 'AURORA'

International Nuclear Information System (INIS)

Takayama, Takeshi; Yano, Takashi; Sasaki, Yasushi; Yasumitsu, Naoki

1991-01-01

A half-integer-resonance injection method is introduced for a superconducting SR-ring of 1 m orbit diameter, which is made of a weak focussing single-body magnet. The present method makes it possible to inject an electron beam of an energy of as high as 150 MeV into the ring of a magnetic field strength of 1 T. Several new injection devices are introduced in order to guide the beam under the strong magnetic fringing field, and to excite the half-integer-resonance. The field index of 0.73 is selected for the half-integer-resonance injection. The field index of 0.35 at the maximum magnetic field strength of 4.3 T is to get a sufficiently long quantum lifetime. A new device named resonance jumper is used to pass quickly several resonances of betatron motion without beam loss. The resonances occur when the magnetic field is ramped up and the field index decreases from 0.73 to 0.35. The injection devices except the inflector are air-core magnets in order to work in the strong magnetic field. In November of 1989, the beam was successfully injected and stored. The injection devices and the half-integer-resonance injection method were established. (author)

GH and IGF-I induction by passive immunization of rainbow trout Oncorhynchus mykiss Walbaum using a somatostatin 14 antibody

Science.gov (United States)

Inhibition of the growth axis by somatostatin was studied in juvenile rainbow trout using passive immunization with a previously isolated somatostatin antibody (antiSS-14). Upon subcutaneously injection of laying hens (Gallus domesticus) with conjugated somatostatin-14 (SS-14), the antiSS-14 was iso...

Synthesis and characterization of ZnSe:Fe/ZnSe core/shell nanocrystals

Energy Technology Data Exchange (ETDEWEB)

Yang, Lin; Zhu, Jianguo, E-mail: yanglin_1028@163.com; Xiao, Dingquan

2014-04-15

High-quality ZnSe:Fe/ZnSe core/shell nanocrystals were prepared via a hydrothermal microemulsion technique. Effective surface passivation of monodisperse ZnSe:Fe nanocrystals is achieved by overcoating them with a ZnSe shell. The samples were characterized by means of XRD, EDX, TEM, PSD, XPS, photoluminescence, and Raman spectrum. The results show that the as-synthesized nanocrystals are cubic zinc blende ZnSe structure with high purity and the average particle size of ZnSe:Fe/ZnSe core/shell nanocrystal is larger than that of ZnSe:Fe core. The growth of ZnSe shell causes a small red shift in PL spectra, and then the PL quantum yield (QY) increases from 16% before shell growth to the maximum of 37% after increasing shell thickness up to 1.2 monolayers (ML). Moreover, both transverse optic (TO) and longitudinal optic (LO) phonon modes of ZnSe are shifted toward lower frequency as compared with the reported ones. -- Highlights: • ZnSe:Fe/ZnSe core/shell QDs were prepared by a hydrothermal microemulsion method. • ZnSe shell efficiently passivates surface defects by serving as a physical barrier. • The particle size and PL properties can be turned with the growth of ZnSe shell. • The luminescence efficiency and stability of QDs could be improved in this manner.

Welding lines formation in holes obtained by low pressure injection molding of ceramic parts

Directory of Open Access Journals (Sweden)

C. A. Costa

Full Text Available Abstract This work presents a study to evaluate the process of producing internal holes in ceramic disks produced by low pressure injection molding (LPIM process. Two process conditions defined as pre-injection and post-injection were used to test the proposition. In the first one the pin cores that produce the holes were positioned in the cavity before the injection of the feedstock; and in the second one, the pin cores were positioned in the cavity, just after the feeding phase of the injection mold. An experimental injection mold designed and manufactured to test both processes was developed to produce ceramic disk with Ø 50 x 2 mm with four holes of Ø 5 mm, equally and radially distributed through the disk. The feedstock was composed of 86 wt% alumina (Al2O3 and 14 wt% organic vehicle based on paraffin wax. Heating and cooling systems controlled by a data acquisition system were included in the mold. The results showed that there were no welding lines with the post-injection process, proving to be an option for creating holes in the ceramic parts produced by LPIM. It was observed that best results were obtained at 58 °C mold temperature. The pins extraction temperature was about 45 °C, and the injection pressure was 170 kPa.

SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems

International Nuclear Information System (INIS)

Carsten, Pasler

2007-01-01

The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand

Research on removing reservoir core water sensitivity using the method of ultrasound-chemical agent for enhanced oil recovery.

Science.gov (United States)

Wang, Zhenjun; Huang, Jiehao

2018-04-01

The phenomenon of water sensitivity often occurs in the oil reservoir core during the process of crude oil production, which seriously affects the efficiency of oil extraction. In recent years, near-well ultrasonic processing technology attaches more attention due to its safety and energy efficient. In this paper, the comparison of removing core water sensitivity by ultrasonic wave, chemical injection and ultrasound-chemical combination technique are investigated through experiments. Results show that: lower ultrasonic frequency and higher power can improve the efficiency of core water sensitivity removal; the effects of removing core water sensitivity under ultrasonic treatment get better with increase of core initial permeability; the effect of removing core water sensitivity using ultrasonic treatment won't get better over time. Ultrasonic treatment time should be controlled in a reasonable range; the effect of removing core water sensitivity using chemical agent alone is slightly better than that using ultrasonic treatment, however, chemical injection could be replaced by ultrasonic treatment for removing core water sensitivity from the viewpoint of oil reservoir protection and the sustainable development of oil field; ultrasound-chemical combination technique has the best effect for water sensitivity removal than using ultrasonic treatment or chemical injection alone. Copyright © 2017 Elsevier B.V. All rights reserved.

Experimental study of solvent-based emulsion injection to enhance heavy oil recovery in Alaska North Slope area

Energy Technology Data Exchange (ETDEWEB)

Qiu, F.; Mamora, D. [Texas A and M Univ., College Station, TX (United States)

2010-07-01

This study examined the feasibility of using a chemical enhanced oil recovery method to overcome some of the technical challenges associated with thermal recovery in the Alaska North Slope (ANS). This paper described the second stage research of an experimental study on nano-particle and surfactant-stabilized solvent-based emulsions for the ANS area. Four successful core flood experiments were performed using heavy ANS oil. The runs included water flooding followed by emulsion flooding; and pure emulsion injection core flooding. The injection rate and core flooding temperature remained constant and only 1 PV micro-emulsion was injected after breakthrough under water flooding or emulsion flooding. Oil recovery increased by 26.4 percent from 56.2 percent original oil in place (OOIP) with waterflooding to 82.6 percent OOIP with injection of emulsion following water flooding. Oil recovery was slightly higher with pure emulsion flooding, at 85.8 percent OOIP. The study showed that low permeability generally resulted in a higher shear rate, which is favourable for in-situ emulsification and higher displacement efficiency. 11 refs., 4 tabs., 20 figs.

Development of small, fast reactor core designs using lead-based coolant

International Nuclear Information System (INIS)

Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

1999-01-01

A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

Interpretation of risk significance of passive component aging using probabilistic structural analysis

International Nuclear Information System (INIS)

Phillips, J.H.; Atwood, C.L.

1993-01-01

The probabilistic risk assessments (PRAs) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. Except as initiating events, the possible failure of passive components is given little consideration. The NRC is sponsoring a project at INEL to investigate the risk significance of passive components as they age. For this project, we developed a technique to calculate the failure probability of passive components over time, and demonstrated the technique by applying it to a weld in the auxiliary feedwater (AFW) system. A decreasing yearly rupture rate for this weld was calculated instead of the increasing rupture rate trend one might expect. We attribute this result to infant mortality; that is, most of those initial flaws that will eventually lead to rupture will do so early in life. This means that although each weld in a population may be wearing out, the population as a whole can exhibit a decreasing rupture rate. This observation has implications for passive components in commercial nuclear plants and other facilities where aging is a concern. For the population of passive components that exhibit a decreasing failure rate, risk increase is not a concern. The next step of the work is to identify the attributes that contribute to this decreasing rate, and to determine any attributes that would contribute to an increasing failure rate and thus to an increased risk

Injection characteristics of dimethyl ether

Energy Technology Data Exchange (ETDEWEB)

Glensvig, M.

1996-09-01

Dimethyl ether (DME) has proved to be a new ultra-clean alternative fuel for diesel engines. Engine tests have shown considerably lower NO{sub x} emissions, no particle emissions and lower noise compared to that obtained from normal diesel engine operation. DME also has demonstrated favorable response to Exhaust Gas Recirculation (EGR). The purpose of this investigation was to achieve a better understanding of the fundamental spray behavior of DME. Fundamental spray behaviour was characterized by fuel spray penetration and angle, atomization and droplet size and evaporation. The influence of fuel characteristics, nozzle geometry and ambient pressure on the DME and diesel spray behavior was investigated. Fuel was injected into an unheated injection chamber with a ambient pressure of 15 bar and 25 bar, respectively, giving a simplified simulation of the environment in an operating engine. Two nozzles were studied: a single hole nozzle and a pintle nozzle. A conventional fuel injection system was used for both nozzles. Injection parameters of RPM, throttle position, fuel line length and chamber environment were held constant for both nozzles. The sprays were visualized using schlieren and high speed photography. Results show that the general appearance of the DME spray is similar to that of diesel spray. The core of the DME spray seems less dense and the spray tip less sharp compared to diesel spray, indicating smaller droplets with a lower momentum in the core of the DME spray. Schlieren film shows that with both DME and diesel fuel, the spray tip only consists of liquid and that evaporation occurs after a brief time interval. Penetration of DME is about one third that of diesel using the pintle nozzle. Also, the spray angle is considerably larger for the DME spray compared to the diesel spray. A comparatively smaller difference in penetration is observed using the hole nozzle. Differences in penetration for the hole nozzle are within the limit of the penetration

The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

International Nuclear Information System (INIS)

Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

1987-03-01

A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

Evaluation of remote delivery of Passive Integrated Transponder (PIT technology to mark large mammals.

Directory of Open Access Journals (Sweden)

W David Walter

Full Text Available Methods to individually mark and identify free-ranging wildlife without trapping and handling would be useful for a variety of research and management purposes. The use of Passive Integrated Transponder technology could be an efficient method for collecting data for mark-recapture analysis and other strategies for assessing characteristics about populations of various wildlife species. Passive Integrated Transponder tags (PIT have unique numbered frequencies and have been used to successfully mark and identify mammals. We tested for successful injection of PIT and subsequent functioning of PIT into gelatin blocks using 4 variations of a prototype dart. We then selected the prototype dart that resulted in the least depth of penetration in the gelatin block to assess the ability of PIT to be successfully implanted into muscle tissue of white-tailed deer (Odocoileus virginianus post-mortem and long-term in live, captive Rocky Mountain elk (Cervus elaphus. The prototype dart with a 12.7 mm (0.5 inch needle length and no powder charge resulted in the shallowest mean (± SD penetration depth into gelatin blocks of 27.0 mm (± 5.6 mm with 2.0 psi setting on the Dan-Inject CO(2-pressured rifle. Eighty percent of PIT were successfully injected in the muscle mass of white-tailed deer post-mortem with a mean (± SD penetration depth of 22.2 mm (± 3.8 mm; n = 6. We injected PIT successfully into 13 live, captive elk by remote delivery at about 20 m that remained functional for 7 months. We successfully demonstrated that PIT could be remotely delivered in darts into muscle mass of large mammals and remain functional for >6 months. Although further research is warranted to fully develop the technique, remote delivery of PIT technology to large mammals is possible using prototype implant darts.

Feasibility and application on steam injector for next-generation reactor

International Nuclear Information System (INIS)

Narabayashi, Tadashi; Ishiyama, Takenori; Miyano, Hiroshi; Nei, Hiromichi; Shioiri, Akio

1991-01-01

A feasibility study has been conducted on steam injector for a next generation reactor. The steam injector is a simple, compact passive device for water injection, such as Passive Core Injection System (PCIS) of Passive Containment Cooling System (PCCS), because of easy start-up without an AC power. An analysis model for a steam injector characteristics has been developed, and investigated with a visualized fundamental test for a two-stage Steam Injector System (SIS) for PCIS and a one-stage low pressure SIS for PCCS. The test results showed good agreement with the analysis results. The analysis and the test results showed the SIS could work over a very wide range of the steam pressure, and is applicable for PCIS or PCCS in the next generation reactors. (author)

Passive film growth on carbon steel and its nanoscale features at various passivating potentials

International Nuclear Information System (INIS)

Li, Yuan; Cheng, Y. Frank

2017-01-01

Highlights: • Imaged the topography of passivated steel at various film-forming potentials. • Characterized the nanoscale features of passive films. • Determined the composition of passive films formed at various potentials. - Abstract: In this work, the passivation and topographic sub-structure of passive films on a carbon steel in a carbonate/bicarbonate solution was characterized by electrochemical measurements, atomic force microscopy and X-ray photoelectron spectroscopy. When passivating at a potential near the active-passive transition, the film contains the mixture of Fe_3O_4, Fe_2O_3 and FeOOH, with numerous nanoscale features. As the film-forming potential shifts positively, the passive film becomes more compact and the nanoscale features disappear. When the film is formed at a passive potential where the oxygen evolution is enabled, the content of FeOOH in the film increases, resulting in an amorphous topography and reduced corrosion resistance.

Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

International Nuclear Information System (INIS)

Knebel, J.U.; Mueller, U.

1997-01-01

This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

Energy Technology Data Exchange (ETDEWEB)

Knebel, J.U.; Mueller, U. [Forschungszentrum Karlsruhe - Technik und Umwelt Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF), Karlsruhe (Germany)

1997-12-31

This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

Passivation of fluorinated activated charcoal

International Nuclear Information System (INIS)

Del Cul, G.D.; Trowbridge, L.D.; Simmons, D.W.; Williams, D.F.; Toth, L.M.

1997-10-01

The Molten Salt Reactor Experiment (MSRE), at the Oak Ridge National Laboratory has been shut down since 1969 when the fuel salt was drained from the core into two Hastelloy N tanks at the reactor site. In 1995, a multiyear project was launched to remediate the potentially hazardous conditions generated by the movement of fissile material and reactive gases from the storage tanks into the piping system and an auxiliary charcoal bed (ACB). The top 12 in. of the ACB is known by gamma scan and thermal analysis to contain about 2.6 kg U-233. According to the laboratory tests, a few feet of fluorinated charcoal are believed to extend beyond the uranium front. The remainder of the ACB should consist of unreacted charcoal. Fluorinated charcoal, when subjected to rapid heating, can decompose generating gaseous products. Under confined conditions, the sudden exothermic decomposition can produce high temperatures and pressures of near-explosive characteristics. Since it will be necessary to drill and tap the ACB to allow installation of piping and instrumentation for remediation and recovery activities, it is necessary to chemically convert the reactive fluorinated charcoal into a more stable material. Ammonia can be administered to the ACB as a volatile denaturing agent that results in the conversion of the C x F to carbon and ammonium fluoride, NH 4 F. The charcoal laden with NH 4 F can then be heated without risking any sudden decomposition. The only consequence of heating the treated material will be the volatilization of NH 4 F as a mixture of NH 3 and HF, which would primarily recombine as NH 4 F on surfaces below 200 C. The planned scheme for the ACB denaturing is to flow diluted ammonia gas in steps of increasing NH 3 concentration, 2% to 50%, followed by the injection of pure ammonia. This report summarizes the planned passivation treatment scheme to stabilize the ACB and remove the potential hazards. It also includes basic information, results of laboratory tests

Curcumin exerts neuroprotective effects against homocysteine intracerebroventricular injection-induced cognitive impairment and oxidative stress in rat brain.

Science.gov (United States)

Ataie, Amin; Sabetkasaei, Masoumeh; Haghparast, Abbas; Moghaddam, Akbar Hajizadeh; Ataee, Ramin; Moghaddam, Shiva Nasiraei

2010-08-01

Aging is the major risk factor for neurodegenerative diseases and oxidative stress and is involved in their pathophysiology. Oxidative stress can induce neuronal damage and modulate intracellular signaling, ultimately leading to neuronal death by apoptosis or necrosis. In this study we investigated the neuroprotective properties of the natural polyphenolic antioxidant compound, curcumin, against homocysteine (Hcy) neurotoxicity. Curcumin (5, 15, or 45 mg/kg) was injected intraperitoneally once daily for a period of 10 days beginning 5 days prior to Hcy (0.2 micromol/microl) intracerebroventricular injection in rats. Biochemical and behavioral studies, including passive avoidance learning and locomotor activity tests, were evaluated 24 hours after the last injection of curcumin or vehicle. Results indicated that Hcy induces lipid peroxidation and increases malondialdehyde (MDA) and superoxide anion (SOA) levels in whole rat brain. In addition, Hcy impaired memory retention in the passive avoidance learning test. However, curcumin treatment significantly decreased MDA and SOA levels and improved learning and memory in rats. These results suggest that Hcy may induce lipid peroxidation in rat brain and that polyphenol treatment (curcumin) improves learning and memory deficits by protecting the nervous system against oxidative stress.

Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Feldman, E.M.

1977-03-01

Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-4 was conducted from initial conditions of 2266 psia and 543 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

ZnSe passivation layer for the efficiency enhancement of CuInS2 quantum dots sensitized solar cells

International Nuclear Information System (INIS)

Peng, Zhuoyin; Liu, Yueli; Zhao, Yinghan; Chen, Keqiang; Cheng, Yuqing; Kovalev, Valery; Chen, Wen

2014-01-01

Highlights: • ZnSe is employed as passivation layer in CuInS 2 quantum dots sensitized solar cells. • Slight red-shift has been occurred in UV–vis absorption spectra with ZnSe coating. • CuInS 2 based solar cells coated by ZnSe have better efficiency than that of ZnS. • Higher rate of charge transport can be produced after coating with ZnSe. -- Abstract: The effect of ZnSe passivation layer is investigated in the CuInS 2 quantum dot sensitized solar cells, which is used to improve the photovoltaic performance. The CuInS 2 quantum dot sensitized TiO 2 photo-anodes are prepared by assembly linking technique, and then deposited by the ZnSe passivation layer using the successive ionic layer absorption and reaction technique. The optical absorption edge and photoluminescence peak have slightly red-shifted after the passivation layer coating. Under solar light illumination, the ZnSe passivation layer based CuInS 2 quantum dot sensitized solar cells have the higher photovoltaic efficiency of 0.95% and incident photon conversion efficiency response than that of pure CuInS 2 based solar cells and ZnS passivation layer based solar cells, as the electron injection rate becomes faster after coating with ZnSe passivation layer

Including 10-Gigabit-capable Passive Optical Network under End-to-End Generalized Multi-Protocol Label Switching Provisioned Quality of Service

DEFF Research Database (Denmark)

Brewka, Lukasz Jerzy; Gavler, Anders; Wessing, Henrik

2012-01-01

of the network where quality of service signaling is bridged. This article proposes strategies for generalized multi-protocol label switching control over next emerging passive optical network standard, i.e., the 10-gigabit-capable passive optical network. Node management and resource allocation approaches...... are discussed, and possible issues are raised. The analysis shows that consideration of a 10-gigabit-capable passive optical network as a generalized multi-protocol label switching controlled domain is valid and may advance end-to-end quality of service provisioning for passive optical network based customers.......End-to-end quality of service provisioning is still a challenging task despite many years of research and development in this area. Considering a generalized multi-protocol label switching based core/metro network and resource reservation protocol capable home gateways, it is the access part...

Passive film growth on carbon steel and its nanoscale features at various passivating potentials

Energy Technology Data Exchange (ETDEWEB)

Li, Yuan; Cheng, Y. Frank, E-mail: fcheng@ucalgary.ca

2017-02-28

Highlights: • Imaged the topography of passivated steel at various film-forming potentials. • Characterized the nanoscale features of passive films. • Determined the composition of passive films formed at various potentials. - Abstract: In this work, the passivation and topographic sub-structure of passive films on a carbon steel in a carbonate/bicarbonate solution was characterized by electrochemical measurements, atomic force microscopy and X-ray photoelectron spectroscopy. When passivating at a potential near the active-passive transition, the film contains the mixture of Fe{sub 3}O{sub 4}, Fe{sub 2}O{sub 3} and FeOOH, with numerous nanoscale features. As the film-forming potential shifts positively, the passive film becomes more compact and the nanoscale features disappear. When the film is formed at a passive potential where the oxygen evolution is enabled, the content of FeOOH in the film increases, resulting in an amorphous topography and reduced corrosion resistance.

Passivity-based design of robust passive damping for LCL-filtered voltage source converters

DEFF Research Database (Denmark)

Wang, Xiongfei; Blaabjerg, Frede; Loh, Poh Chiang

2015-01-01

Passive damping is proven as a robust stabilizing technique for LCL-filtered voltage source converters. However, conventional design methods of passive dampers are based on the passive components only, while the inherent damping effect of time delay in the digital control system is overlooked....... In this paper, a frequency-domain passivity-based design approach is proposed, where the passive dampers are designed to eliminate the negative real part of the converter output admittance with closed-loop current control, rather than shaping the LCL-filter itself. Thus, the influence of time delay...... in the current control is included, which allows a relaxed design of the passive damper with the reduced power loss and improved stability robustness against grid parameters variations. Design procedures of two commonly used passive dampers with LCL-filtered VSCs are illustrated. Experimental results validate...

Optimization of Wireless Power Transfer Systems Enhanced by Passive Elements and Metasurfaces

Science.gov (United States)

Lang, Hans-Dieter; Sarris, Costas D.

2017-10-01

This paper presents a rigorous optimization technique for wireless power transfer (WPT) systems enhanced by passive elements, ranging from simple reflectors and intermedi- ate relays all the way to general electromagnetic guiding and focusing structures, such as metasurfaces and metamaterials. At its core is a convex semidefinite relaxation formulation of the otherwise nonconvex optimization problem, of which tightness and optimality can be confirmed by a simple test of its solutions. The resulting method is rigorous, versatile, and general -- it does not rely on any assumptions. As shown in various examples, it is able to efficiently and reliably optimize such WPT systems in order to find their physical limitations on performance, optimal operating parameters and inspect their working principles, even for a large number of active transmitters and passive elements.

Muscle structure and stiffness assessment after botulinum toxin type A injection. A systematic review.

Science.gov (United States)

Mathevon, L; Michel, F; Decavel, P; Fernandez, B; Parratte, B; Calmels, P

2015-12-01

Botulinum toxin type A manages spasticity disorders in neurological central diseases. Some studies have reported that it might induce muscle changes. We present a literature review abiding by the PRISMA statement guidelines. The purpose was to explore the structural and passive biomechanical muscle properties after botulinum toxin type A injections in healthy and spastic limb muscles, on animals and humans, as well as methods for evaluating these properties. We searched the PubMed and Cochrane Library databases using the following keywords: "Botulinum toxin" AND ("muscle structure" OR "muscle atrophy") and, "Botulinum toxin" AND "muscle elasticity". From the 228 initially identified articles, 21 articles were included. Histological analyses were performed, especially on animals. A neurogenic atrophy systematically occurred. In humans, one year after a single injection, the histological recovery remained incomplete. Furthermore, 2D ultrasound analyses showed a reduction of the gastrocnemius thickness and pennation angle. MRI volumetric analysis evidenced muscular atrophy six months or one year after a single injection. Passive muscle stiffness depends on these structural changes. On the short term, the biomechanical analysis showed an elastic modulus increase in animals whereas no change was recorded in humans. On the short term, ultrasound elastography imaging showed a decreased elastic modulus. To date, few data are available, but all show a structural and mechanical muscle impact post injections, specifically muscle atrophy which can linger over time. Further studies are necessary to validate this element, and the possibility of change must be taken into account particularly with repeated injections. Thus, in clinical practice, 2D ultrasound and ultrasound elastography are two non-invasive techniques that will help physicians to develop an efficient long term monitoring. Copyright © 2015 Elsevier Masson SAS. All rights reserved.

Passive Resonant Bidirectional Converter with Galvanic Barrier

Science.gov (United States)

Rosenblad, Nathan S. (Inventor)

2014-01-01

A passive resonant bidirectional converter system that transports energy across a galvanic barrier includes a converter using at least first and second converter sections, each section including a pair of transfer terminals, a center tapped winding; a chopper circuit interconnected between the center tapped winding and one of the transfer terminals; an inductance feed winding interconnected between the other of the transfer terminals and the center tap and a resonant tank circuit including at least the inductance of the center tap winding and the parasitic capacitance of the chopper circuit for operating the converter section at resonance; the center tapped windings of the first and second converter sections being disposed on a first common winding core and the inductance feed windings of the first and second converter sections being disposed on a second common winding core for automatically synchronizing the resonant oscillation of the first and second converter sections and transferring energy between the converter sections until the voltage across the pairs of transfer terminals achieves the turns ratio of the center tapped windings.

Emergency core cooling systems

International Nuclear Information System (INIS)

Kubokoya, Takashi; Okataku, Yasukuni.

1984-01-01

Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

Passivity and Evolutionary Game Dynamics

KAUST Repository

Park, Shinkyu; Shamma, Jeff S.; Martins, Nuno C.

2018-01-01

This paper investigates an energy conservation and dissipation -- passivity -- aspect of dynamic models in evolutionary game theory. We define a notion of passivity using the state-space representation of the models, and we devise systematic methods to examine passivity and to identify properties of passive dynamic models. Based on the methods, we describe how passivity is connected to stability in population games and illustrate stability of passive dynamic models using numerical simulations.

Passivity and Evolutionary Game Dynamics

KAUST Repository

Park, Shinkyu

2018-03-21

This paper investigates an energy conservation and dissipation -- passivity -- aspect of dynamic models in evolutionary game theory. We define a notion of passivity using the state-space representation of the models, and we devise systematic methods to examine passivity and to identify properties of passive dynamic models. Based on the methods, we describe how passivity is connected to stability in population games and illustrate stability of passive dynamic models using numerical simulations.

Perspectives on Severe Accident Management by Depressurization and External Water Injection under Extended SBO Conditions

International Nuclear Information System (INIS)

Seol, Wookcheol; Park, Jongwoon

2014-01-01

Three major issues of severe accident management guideline (SAMG) after this sort of extended SBO would be depressurization of the primary system, external water injection and hydrogen management inside a containment. Under this situation, typical SAM actions would be depressurization and external water delivery into the core. However, limited amount of external water would necessitate optimization between core cooling, containment integrity and fission product removal. In this paper, effects of SAM actions such as depressurization and external water injection on the reactor and containment conditions after extended SBO are analyzed using MAAP4 code. Positive and negative aspects are discussed with respect to core cooling and fission product retention inside a primary system. Conclusions are made as following: Firstly, early depressurization action itself has two-faces: positive with respect to delay of the reactor vessel failure but negative with respect to the containment failure and fission product retention inside the primary system. Secondly, in order to prevent containment overpressure failure after external water injection, re-closing of PORV later should be considered in SAM, which has never been considered in the previous SAMG. Finally, in case of external water injection, the flow rate should be optimized considering not only the cooling effect but also the long term fission product retention inside the primary system

Performance of the prism reactor's passive decay heat removal system

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

Laboratory scale micro-seismic monitoring of rock faulting and injection-induced fault reactivation

Science.gov (United States)

Sarout, J.; Dautriat, J.; Esteban, L.; Lumley, D. E.; King, A.

2017-12-01

The South West Hub CCS project in Western Australia aims to evaluate the feasibility and impact of geosequestration of CO2 in the Lesueur sandstone formation. Part of this evaluation focuses on the feasibility and design of a robust passive seismic monitoring array. Micro-seismicity monitoring can be used to image the injected CO2plume, or any geomechanical fracture/fault activity; and thus serve as an early warning system by measuring low-level (unfelt) seismicity that may precede potentially larger (felt) earthquakes. This paper describes laboratory deformation experiments replicating typical field scenarios of fluid injection in faulted reservoirs. Two pairs of cylindrical core specimens were recovered from the Harvey-1 well at depths of 1924 m and 2508 m. In each specimen a fault is first generated at the in situ stress, pore pressure and temperature by increasing the vertical stress beyond the peak in a triaxial stress vessel at CSIRO's Geomechanics & Geophysics Lab. The faulted specimen is then stabilized by decreasing the vertical stress. The freshly formed fault is subsequently reactivated by brine injection and increase of the pore pressure until slip occurs again. This second slip event is then controlled in displacement and allowed to develop for a few millimeters. The micro-seismic (MS) response of the rock during the initial fracturing and subsequent reactivation is monitored using an array of 16 ultrasonic sensors attached to the specimen's surface. The recorded MS events are relocated in space and time, and correlate well with the 3D X-ray CT images of the specimen obtained post-mortem. The time evolution of the structural changes induced within the triaxial stress vessel is therefore reliably inferred. The recorded MS activity shows that, as expected, the increase of the vertical stress beyond the peak led to an inclined shear fault. The injection of fluid and the resulting increase in pore pressure led first to a reactivation of the pre

Biodistribution of radiolabelled human dendritic cells injected by various routes

International Nuclear Information System (INIS)

Quillien, Veronique; Moisan, Annick; Carsin, Andre; Lesimple, Thierry; Lefeuvre, Claudia; Bertho, Nicolas; Devillers, Anne; Toujas, Louis; Adamski, Henri; Leberre, Claudine

2005-01-01

The purpose of this study was to investigate the biodistribution of mature dendritic cells (DCs) injected by various routes, during a cell therapy protocol. In the context of a vaccine therapy protocol for melanoma, DCs matured with Ribomunyl and interferon-gamma were labelled with 111 In-oxine and injected into eight patients along various routes: afferent lymphatic vessel (IL) (4 times), lymph node (IN) (5 times) and intradermally (ID) (6 times). Scintigraphic investigations showed that the IL route allowed localisation of 80% of injected radioactivity in eight to ten nodes. In three cases of IN injection, the entire radioactivity stagnated in the injected nodes, while in two cases, migration to adjacent nodes was observed. This migration was detected rapidly after injection, as with IL injections, suggesting that passive transport occurred along the physiological lymphatic pathways. In two of the six ID injections, 1-2% of injected radioactivity reached a proximal lymph node. Migration was detectable in the first hour, but increased considerably after 24 h, suggesting an active migration mechanism. In both of the aforementioned cases, DCs were strongly CCR7-positive, but this feature was not a sufficient condition for effective migration. In comparison with DCs matured with TNF-α, IL-1β, IL-6 and PGE2, our DCs showed a weaker in vitro migratory response to CCL21, despite comparable CCR7 expression, and higher allostimulatory and TH1 polarisation capacities. The IL route allowed reproducible administration of specified numbers of DCs. The IN route sometimes yielded fairly similar results, but not reproducibly. Lastly, we showed that DCs matured without PGE2 that have in vitro TH1 polarisation capacities can migrate to lymph nodes after ID injection. (orig.)

Evaluation on driving force of natural circulation in downcomer for passive residual heat removal system in JAERI passive safety reactor JPSR

International Nuclear Information System (INIS)

Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

1997-01-01

The driving-force of the natural circulation in the residual heat removal (RHR) system for the JPSR (JAERI Passive Safety Reactor) is given as a gravity force of the density difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The amount of density difference and time to achieve the enough density difference for the RHR system change directly dependent on the thermal fluid flow pattern in downcomer of annulus flow pass. The purposes of the present study are to investigate the possibilities of the followings by evaluating the three-dimensional thermal fluid flow in downcomer by numerical analysis using the STREAM code; 1) promotion of making the flow pattern uniform in downcomer by installing a baffle, 2) achievement of an enough driving-force of the natural circulation, 3) validity of one-point assumption, that is, complete mixing down-flow assumption for the three-dimensional thermal fluid flow in downcomer to evaluate the function of the passive RHR system. The following conclusions were obtained: (1) The effect of baffle on the thermal fluid flow and driving-force is little, (2) The driving-force required for natural circulation cooling can be obtained in wide range of inlet velocity even if the flow is multi-dimensional, (3) Both in initial transient stage and in steady-state, the one-point assumption can be applied to evaluate the driving-force of natural circulation in the passive RHR system. (author)

Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-03-01

Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-3 was conducted from initial conditions of 2263 psia and 545 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

Alpha-MSH decreases core and brain temperature during global cerebral ischemia in rats

DEFF Research Database (Denmark)

Spulber, S.; Moldovan, Mihai; Oprica, M.

2005-01-01

-vessel occlusion forebrain ischemia on core temperature (CT) and brain temperature (BT), respectively. After 10 min cerebral ischemia, BT was lower in alpha-MSH- than in saline-injected animals. After 10 min reperfusion, both CT and BT were lower than the corresponding pre-ischemic levels after injection of alpha...

ACTIVE GALACTIC NUCLEUS FEEDBACK AND ENTROPY INJECTION IN GALAXY CLUSTER CORES

International Nuclear Information System (INIS)

Chaudhuri, Anya; Majumdar, Subhabrata; Nath, Biman B.

2013-01-01

We make the first estimate of non-gravitational energy profiles in galaxy cluster cores (and beyond) based on observational data. Comparing the observed entropy profiles within r 500 , from the Representative XMM-Newton Cluster Structure Survey to simulated base entropy profiles without feedback from both adaptive mesh refinement (AMR) and smoothed particle hydrodynamic (SPH) non-radiative simulations, we estimate the amount of additional non-gravitational energy, E ICM , contained in the intracluster medium (ICM), as well as the total energy feedback, E Feedback , from active galactic nuclei (AGNs; the central AGNs in most cases) into the clusters. The total feedback energy scales with the mean spectroscopic temperature as E Feedback ∝T sp 2.52±0.08 and E Feedback ∝T sp 2.17±0.11 for the SPH and AMR baseline profiles. The mean non-gravitational energy per particle within r 500 remaining in the ICM after energy lost during cooling is ε ICM = 2.8 ± 0.8 keV for the SPH theoretical relation and ε ICM = 1.7 ± 0.9 keV for the AMR theoretical relation. We use the NRAO/VLA Sky Survey source catalog to determine the radio luminosity, L R , at 1.4 GHz of the central source(s) of our sample. For T sp > 3 keV, the E Feedback correlates with L R , although with different normalization for cool-core and non-cool-core clusters. We show that AGNs could provide a significant portion of the feedback

ZnSe passivation layer for the efficiency enhancement of CuInS{sub 2} quantum dots sensitized solar cells

Energy Technology Data Exchange (ETDEWEB)

Peng, Zhuoyin; Liu, Yueli; Zhao, Yinghan; Chen, Keqiang; Cheng, Yuqing [State Key Laboratory of Advanced Technology for Materials Synthesis and Processing, School of Materials Science and Engineering, Wuhan University of Technology, Wuhan 430070 (China); Kovalev, Valery [Department of Mechanics and Mathematics, Moscow State University named after M.V. Lomonosov, Leninskie Gory 1, 119992 Moscow (Russian Federation); Chen, Wen, E-mail: chenw@whut.edu.cn [State Key Laboratory of Advanced Technology for Materials Synthesis and Processing, School of Materials Science and Engineering, Wuhan University of Technology, Wuhan 430070 (China)

2014-02-25

Highlights: • ZnSe is employed as passivation layer in CuInS{sub 2} quantum dots sensitized solar cells. • Slight red-shift has been occurred in UV–vis absorption spectra with ZnSe coating. • CuInS{sub 2} based solar cells coated by ZnSe have better efficiency than that of ZnS. • Higher rate of charge transport can be produced after coating with ZnSe. -- Abstract: The effect of ZnSe passivation layer is investigated in the CuInS{sub 2} quantum dot sensitized solar cells, which is used to improve the photovoltaic performance. The CuInS{sub 2} quantum dot sensitized TiO{sub 2} photo-anodes are prepared by assembly linking technique, and then deposited by the ZnSe passivation layer using the successive ionic layer absorption and reaction technique. The optical absorption edge and photoluminescence peak have slightly red-shifted after the passivation layer coating. Under solar light illumination, the ZnSe passivation layer based CuInS{sub 2} quantum dot sensitized solar cells have the higher photovoltaic efficiency of 0.95% and incident photon conversion efficiency response than that of pure CuInS{sub 2} based solar cells and ZnS passivation layer based solar cells, as the electron injection rate becomes faster after coating with ZnSe passivation layer.

Resolution of issues related to alternative RCS injection in the absence of containment sump recirculation

International Nuclear Information System (INIS)

Charles L Kling; Stephen S Barshay; Mathew C Jacob; Michael J Friedman

2005-01-01

Full text of publication follows: On June 9, 2003 the US NRC issued Bulletin No. 2003-01 that deals with the potential impact of debris blockage on containment sump recirculation at PWRs during a Loss-of-Coolant Accident (LOCA). In response to the bulletin, the Omaha Public Power District (OPPD) is in the process of developing procedural and operational strategies for their Fort Calhoun Station (FCS) to address the issues raised. Westinghouse provided engineering support to OPPD in identifying and resolving issues related to alternative means of supplying safety injection water to the reactor coolant system (RCS) in the absence of containment sump recirculation. Nuclear power plants are designed to protect the core following a LOCA by providing a continuous supply of cooling water to the core. In the long term, the Refueling Water Storage Tank (RWST) inventory will be depleted and core heat removal accomplished via recirculation of water previously injected into the Reactor Coolant System (RCS) and collected in the containment sump. Debris generated within the containment as a result of the impingement of fluid jets in the Zone of Influence (ZOI) of the RCS break and containment wash down may find its way into the containment sump. As the safety injection pumps take suction from the sump, in the recirculation mode of operation, the debris suspended in the sump water could begin to accumulate in the sump screen that is located in the recirculation path. Should sufficient debris accumulate on the sump screen, a flow blockage could potentially develop. This would result in insufficient safety injection pump NPSH, thereby impairing the recirculation mode of injection into RCS. Potential debris blockage and prevention of sump recirculation may be addressed by refilling the RWST with water and injecting this water directly into the core. This paper identifies and attempts to resolve several issues related to this alternative mode of RCS injection. In particular, the

Concept research on general passive system

International Nuclear Information System (INIS)

Han Xu; Yang Yanhua; Zheng Mingguang

2009-01-01

This paper summarized the current passive techniques used in nuclear power plants. Through classification and analysis, the functional characteristics and inherent identification of passive systems were elucidated. By improving and extending the concept of passive system, the general passive concept was proposed, and space and time relativity was discussed and assumption of general passive system were illustrated. The function of idealized general passive system is equivalent with the current passive system, but the design of idealized general passive system is more flexible. (authors)

Preservation of FFTF Data Related to Passive Safety Testing

International Nuclear Information System (INIS)

Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

2010-01-01

One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980's. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: (1) to verify natural circulation as a reliable means to safely remove decay heat, (2) to extend passive safety

Influence of the formation- and passivation rate of boron-oxygen defects for mitigating carrier-induced degradation in silicon within a hydrogen-based model

International Nuclear Information System (INIS)

Hallam, Brett; Abbott, Malcolm; Nampalli, Nitin; Hamer, Phill; Wenham, Stuart

2016-01-01

A three-state model is used to explore the influence of defect formation- and passivation rates of carrier-induced degradation related to boron-oxygen complexes in boron-doped p-type silicon solar cells within a hydrogen-based model. The model highlights that the inability to effectively mitigate carrier-induced degradation at elevated temperatures in previous studies is due to the limited availability of defects for hydrogen passivation, rather than being limited by the defect passivation rate. An acceleration of the defect formation rate is also observed to increase both the effectiveness and speed of carrier-induced degradation mitigation, whereas increases in the passivation rate do not lead to a substantial acceleration of the hydrogen passivation process. For high-throughput mitigation of such carrier-induced degradation on finished solar cell devices, two key factors were found to be required, high-injection conditions (such as by using high intensity illumination) to enable an acceleration of defect formation whilst simultaneously enabling a rapid passivation of the formed defects, and a high temperature to accelerate both defect formation and defect passivation whilst still ensuring an effective mitigation of carrier-induced degradation

Specialists' meeting on passive and active safety features of LMFRs

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-07-01

The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics.

Specialists' meeting on passive and active safety features of LMFRs

International Nuclear Information System (INIS)

1991-01-01

The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics

Surface Passivation in Empirical Tight Binding

OpenAIRE

He, Yu; Tan, Yaohua; Jiang, Zhengping; Povolotskyi, Michael; Klimeck, Gerhard; Kubis, Tillmann

2015-01-01

Empirical Tight Binding (TB) methods are widely used in atomistic device simulations. Existing TB methods to passivate dangling bonds fall into two categories: 1) Method that explicitly includes passivation atoms is limited to passivation with atoms and small molecules only. 2) Method that implicitly incorporates passivation does not distinguish passivation atom types. This work introduces an implicit passivation method that is applicable to any passivation scenario with appropriate parameter...

Experiment of IEA-R1 reactor core cooling by air convection after pool water loss accident

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Baptista Filho, Benedito Dias

2000-01-01

This paper presents a study of a Emergency Core Cooling to be applied to the IEA-R1 reactor. This system must have the characteristics of passive action, with water spraying over the core, and feeding by gravity from elevated reservoirs. In the evaluation, this system must demonstrate that when the reservoirs are emptied, the core cooling must assure to be fulfilled by air natural convection. This work presents the results of temperature distribution in a test section with plates electrically heated simulation the heat generation conditions on the most heated reactor element

Passive Education

Science.gov (United States)

Bojesen, Emile

2018-01-01

This paper does not present an advocacy of a passive education as opposed to an active education nor does it propose that passive education is in any way 'better' or more important than active education. Through readings of Maurice Blanchot, Jacques Derrida and B.S. Johnson, and gentle critiques of Jacques Rancière and John Dewey, passive…

How to arrest a core meltdown accident (doing nothing); Como detener un accidente con fusion de nucleo (sin hacer nada)

Energy Technology Data Exchange (ETDEWEB)

Baron, Jorge H [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

2000-07-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)

Reactivity control in HTR power plants with respect to passive safety system. Summary

Energy Technology Data Exchange (ETDEWEB)

Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

1996-12-01

The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

Operation method and operation control device for emergency core cooling system

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Shoichiro; Takahashi, Toshiyuki; Fujii, Tadashi [Hitachi Ltd., Tokyo (Japan); Mizutani, Akira

1996-05-07

The present invention provides a method of reducing continuous load capacity of an emergency cooling system of a BWR type reactor and a device reducing a rated capacity of an emergency power source facility. Namely, the emergency core cooling system comprises a first cooling system having a plurality of power source systems based on a plurality of emergency power sources and a second cooling system having a remaining heat removing function. In this case, when the first cooling system is operated the manual starting under a predetermined condition that an external power source loss event should occur, a power source division different from the first cooling system shares the operation to operate the secondary cooling system simultaneously. Further, the first cooling system is constituted as a high pressure reactor core water injection system and the second cooling system is constituted as a remaining heat removing system. With such a constitution, a high pressure reactor core water injection system for manual starting and a remaining heat removing system of different power source division can be operated simultaneously before automatic operation of the emergency core cooling system upon loss of external power source of a nuclear power plant. (I.S.)

Hydrogenation of passivated contacts

Energy Technology Data Exchange (ETDEWEB)

Nemeth, William; Yuan, Hao-Chih; LaSalvia, Vincenzo; Stradins, Pauls; Page, Matthew R.

2018-03-06

Methods of hydrogenation of passivated contacts using materials having hydrogen impurities are provided. An example method includes applying, to a passivated contact, a layer of a material, the material containing hydrogen impurities. The method further includes subsequently annealing the material and subsequently removing the material from the passivated contact.

Injection Molding Parameters Calculations by Using Visual Basic (VB) Programming

Science.gov (United States)

Tony, B. Jain A. R.; Karthikeyen, S.; Alex, B. Jeslin A. R.; Hasan, Z. Jahid Ali

2018-03-01

Now a day’s manufacturing industry plays a vital role in production sectors. To fabricate a component lot of design calculation has to be done. There is a chance of human errors occurs during design calculations. The aim of this project is to create a special module using visual basic (VB) programming to calculate injection molding parameters to avoid human errors. To create an injection mold for a spur gear component the following parameters have to be calculated such as Cooling Capacity, Cooling Channel Diameter, and Cooling Channel Length, Runner Length and Runner Diameter, Gate Diameter and Gate Pressure. To calculate the above injection molding parameters a separate module has been created using Visual Basic (VB) Programming to reduce the human errors. The outcome of the module dimensions is the injection molding components such as mold cavity and core design, ejector plate design.

CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN

International Nuclear Information System (INIS)

Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum

1996-06-01

An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author)

CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN

Energy Technology Data Exchange (ETDEWEB)

Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author).

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

2014-01-01

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

Large enhanced dielectric permittivity in polyaniline passivated core-shell nano magnetic iron oxide by plasma polymerization

Energy Technology Data Exchange (ETDEWEB)

Joy, Lija K.; Sooraj, V.; Sethulakshmi, N.; Anantharaman, M. R., E-mail: mraiyer@yahoo.com [Department of Physics, Cochin University of Science and Technology, Cochin-682022, Kerala (India); Sajeev, U. S. [Department of Physics, Government College, Kottayam-686613, Kerala (India); Nair, Swapna S. [Department of Physics, School of Mathematical and Physical Sciences, Central University of Kerala, Kasargode-671123, Kerala (India); Narayanan, T. N. [CSIR-Central Electrochemical Research Institute, Karaikkudi-630006, Tamil Nadu (India); Ajayan, P. M. [Department of Material Science and Nano Engineering, Rice University, 6100 Main Street, Houston, Texas 7700 (United States)

2014-03-24

Commercial samples of Magnetite with size ranging from 25–30 nm were coated with polyaniline by using radio frequency plasma polymerization to achieve a core shell structure of magnetic nanoparticle (core)–Polyaniline (shell). High resolution transmission electron microscopy images confirm the core shell architecture of polyaniline coated iron oxide. The dielectric properties of the material were studied before and after plasma treatment. The polymer coated magnetite particles exhibited a large dielectric permittivity with respect to uncoated samples. The dielectric behavior was modeled using a Maxwell–Wagner capacitor model. A plausible mechanism for the enhancement of dielectric permittivity is proposed.

Passive safety systems and natural circulation in water cooled nuclear power plants

International Nuclear Information System (INIS)

2009-11-01

Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems can eliminate the costs associated with the installation, maintenance, and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

Assessment of the effect of nitrogen gas on passive containment cooling system performance

International Nuclear Information System (INIS)

Ha, Huiun; Suh, Jungsoo

2016-01-01

As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR

Assessment of the effect of nitrogen gas on passive containment cooling system performance

Energy Technology Data Exchange (ETDEWEB)

Ha, Huiun; Suh, Jungsoo [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR.

MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

International Nuclear Information System (INIS)

Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

2003-01-01

OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

Surface Passivation and Antireflection Behavior of ALD on n-Type Silicon for Solar Cells

Directory of Open Access Journals (Sweden)

Ing-Song Yu

2013-01-01

Full Text Available Atomic layer deposition, a method of excellent step coverage and conformal deposition, was used to deposit TiO2 thin films for the surface passivation and antireflection coating of silicon solar cells. TiO2 thin films deposited at different temperatures (200°C, 300°C, 400°C, and 500°C on FZ n-type silicon wafers are in the thickness of 66.4 nm ± 1.1 nm and in the form of self-limiting growth. For the properties of surface passivation, Si surface is effectively passivated by the 200°C deposition TiO2 thin film. Its effective minority carrier lifetime, measured by the photoconductance decay method, is improved 133% at the injection level of  cm−3. Depending on different deposition parameters and annealing processes, we can control the crystallinity of TiO2 and find low-temperature TiO2 phase (anatase better passivation performance than the high-temperature one (rutile, which is consistent with the results of work function measured by Kelvin probe. In addition, TiO2 thin films on polished Si wafer serve as good ARC layers with refractive index between 2.13 and 2.44 at 632.8 nm. Weighted average reflectance at AM1.5G reduces more than half after the deposition of TiO2. Finally, surface passivation and antireflection properties of TiO2 are stable after the cofire process of conventional crystalline Si solar cells.

Field Measurement and Evaluation of the Passive and Active Solar Heating Systems for Residential Building Based on the Qinghai-Tibetan Plateau Case

Directory of Open Access Journals (Sweden)

Zhijian Liu

2017-10-01

Full Text Available Passive and active solar heating systems have drawn much attention and are widely used in residence buildings in the Qinghai-Tibetan plateau due to its high radiation intensity. In fact, there is still lack of quantitative evaluation of the passive and active heating effect, especially for residential building in the Qinghai-Tibetan plateau areas. In this study, three kinds of heating strategies, including reference condition, passive solar heating condition and active solar heating condition, were tested in one demonstration residential building. The hourly air temperatures of each room under different conditions were obtained and analyzed. The results show the indoor air temperature in the living room and bedrooms (core zones was much higher than that of other rooms under both passive and active solar heating conditions. In addition, the heating effect with different strategies for core zones of the building was evaluated by the ratio of indoor and outdoor degree hour, which indicates that solar heating could effectively reduce the traditional energy consumption and improve the indoor thermal environment. The passive solar heating could undertake 49.8% degree hours for heating under an evaluation criterion of 14 °C and the active solar heating could undertake 75% degree hours for heating under evaluation criterion of 18 °C, which indicated that solar heating could effectively reduce the traditional energy consumption and improve the indoor thermal environment in this area. These findings could provide reference for the design and application of solar heating in similar climate areas.

From Dodewaard to a modern economic passive plant-ESBWR

International Nuclear Information System (INIS)

Arnold, H.; Yadigaroglu, G.; Stoop, P.M.; Gonzales, A.; Rao, A.S.

1996-01-01

For over 25 years the Dodewaard nuclear plant has produced electricity with one of the highest reliability of any nuclear power plant in the world. Almost 10 years ago when some designers looked at the features to incorporate in a modern mid-size Boiling Water Reactor (SBWR) design that was developed by an international team and consisted of natural circulation and passive safety injection and decay heat removal. Since the passive decay heat removal was a major new technology area, an extensive test program was developed and conducted - utilizing newly constructed large scale integrated system test facility at the Paul Scherrer Institute, Switzerland. The development of any modern reactor is time consuming and expensive, hence, the design and technology was done as part of an international team effort. This paper provides an overview of the international design and technology effort and discusses how the development costs were minimized through cooperation. The result of the design effort showed that overall plant economics could be improved by a power level increase, supported by some modest design changes. This paper provides the approach and key results of this multi-year international design effort to develop a modern, larger scale, passive plant - the European Simplified Boiling Water Reactor (ESBWR). This effort, in addition to developing an economic nuclear power station, also demonstrates how the benefits of international cooperation can be utilized to keep development costs reasonable. An overview is also provided of the planned research efforts to extend existing technology to the ESBWR. (authors)

Effect of ozone concentration on silicon surface passivation by atomic layer deposited Al2O3

International Nuclear Information System (INIS)

Gastrow, Guillaume von; Li, Shuo; Putkonen, Matti; Laitinen, Mikko; Sajavaara, Timo; Savin, Hele

2015-01-01

Highlights: • The ALD Al 2 O 3 passivation quality can be controlled by the ozone concentration. • Ozone concentration affects the Si/Al 2 O 3 interface charge and defect density. • A surface recombination velocity of 7 cm/s is reached combining ozone and water ALD. • Carbon and hydrogen concentrations correlate with the surface passivation quality. - Abstract: We study the impact of ozone-based Al 2 O 3 Atomic Layer Deposition (ALD) on the surface passivation quality of crystalline silicon. We show that the passivation quality strongly depends on the ozone concentration: the higher ozone concentration results in lower interface defect density and thereby improved passivation. In contrast to previous studies, our results reveal that too high interface hydrogen content can be detrimental to the passivation. The interface hydrogen concentration can be optimized by the ozone-based process; however, the use of pure ozone increases the harmful carbon concentration in the film. Here we demonstrate that low carbon and optimal hydrogen concentration can be achieved by a single process combining the water- and ozone-based reactions. This process results in an interface defect density of 2 × 10 11 eV −1 cm −2 , and maximum surface recombination velocities of 7.1 cm/s and 10 cm/s, after annealing and after an additional firing at 800 °C, respectively. In addition, our results suggest that the effective oxide charge density can be optimized in a simple way by varying the ozone concentration and by injecting water to the ozone process.

Zinc injection helps reduce radiation field buildup in BWRs

International Nuclear Information System (INIS)

Wood, C.

1991-01-01

The injection of zinc into the reactor water of BWRs (Boiling Water Reactors) was a technique developed by General Electric (GE) and the Electric Power Research Institute (EPRI) to control the buildup of radiation fields from cobalt-60 on out-of-core piping. The presence of 5-10ppb zinc in the reactor water reduces the growth of oxide films on stainless steel surfaces, thereby reducing the number of sites available for the incorporation of cobalt; zinc also competes with cobalt for the sites. In September 1990, EPRI organized a workshop at the request of several US utilities to provide a forum to discuss experiences with zinc injection. The meeting focused on six main issues: the effect of zinc on radiation fields in normal water chemistry; the radiation buildup in hydrogen water chemistry, with and without zinc; the effects of zinc-65; the corrosion of fuel cladding and structural materials; the performance of zinc injection and monitoring equipment; and planning for zinc injection. (author)

Passive safety; Passive Sicherheit

Energy Technology Data Exchange (ETDEWEB)

Rueckert, J. [Skoda Auto a.s., Mlada Boleslav (Czech Republic). Interieurentwicklung und Versuche; Hau, M. [Skoda Auto a.s., Mlada Boleslav (Czech Republic). Koordination der Fahrzeugsicherung

2004-05-01

The specifications for passive safety are partly based on the legal requirements for all export markets combined with the strict internal standards of Volkswagen Group. The Euro NCAP tests and their precisely defined testing methods using the new point assessment are very important. (orig.)

Attack methodology Analysis: SQL Injection Attacks and Their Applicability to Control Systems

Energy Technology Data Exchange (ETDEWEB)

Bri Rolston

2005-09-01

Database applications have become a core component in control systems and their associated record keeping utilities. Traditional security models attempt to secure systems by isolating core software components and concentrating security efforts against threats specific to those computers or software components. Database security within control systems follows these models by using generally independent systems that rely on one another for proper functionality. The high level of reliance between the two systems creates an expanded threat surface. To understand the scope of a threat surface, all segments of the control system, with an emphasis on entry points, must be examined. The communication link between data and decision layers is the primary attack surface for SQL injection. This paper facilitates understanding what SQL injection is and why it is a significant threat to control system environments.

Prediction of Gas Injection Performance for Heterogeneous Reservoirs

Energy Technology Data Exchange (ETDEWEB)

Blunt, Martin J.; Orr, Franklin M.

1999-05-17

This report describes research carried out in the Department of Petroleum Engineering at Stanford University from September 1997 - September 1998 under the second year of a three-year grant from the Department of Energy on the "Prediction of Gas Injection Performance for Heterogeneous Reservoirs." The research effort is an integrated study of the factors affecting gas injection, from the pore scale to the field scale, and involves theoretical analysis, laboratory experiments, and numerical simulation. The original proposal described research in four areas: (1) Pore scale modeling of three phase flow in porous media; (2) Laboratory experiments and analysis of factors influencing gas injection performance at the core scale with an emphasis on the fundamentals of three phase flow; (3) Benchmark simulations of gas injection at the field scale; and (4) Development of streamline-based reservoir simulator. Each state of the research is planned to provide input and insight into the next stage, such that at the end we should have an integrated understanding of the key factors affecting field scale displacements.

Passivity-Based Control of a Class of Blondel-Park Transformable Electric Machines

Directory of Open Access Journals (Sweden)

Per J. Nicklasson

1997-10-01

Full Text Available In this paper we study the viability of extending, to the general rotating electric machine's model, the passivity-based controller method that we have developed for induction motors. In this approach the passivity (energy dissipation properties of the motor are taken advantage of at two different levels. First, we prove that the motor model can be decomposed as the feedback interconnection of two passive subsystems, which can essentially be identified with the electrical and mechanical dynamics. Then, we design a torque tracking controller that preserves passivity for the electrical subsystem, and leave the mechanical part as a "passive disturbance". In position or speed control applications this procedure naturally leads to the well known cascaded controller structure which is typically analyzed invoking time-scale separation assumptions. A key feature of the new cascaded control paradigm is that the latter arguments are obviated in the stability analysis. Our objective in this paper is to characterize a class of machines for which such a passivity-based controller solves the output feedback torque tracking problem. Roughly speaking, the class consists of machines whose nonactuated dynamics are well damped and whose electrical and mechanical dynamics can be suitably decoupled via a coordinate transformation. The first condition translates into the requirement of approximate knowledge of the rotor resistances to avoid the need of injecting high gain into the loop. The latter condition is known in the electric machines literature as Blondel-Park transformability, and in practical terms it requires that the air-gap magnetomotive force must be suitably approximated by the first harmonic in its Fourier expansion. These conditions, stemming from the construction of the machine, have a clear physical interpretation in terms of the couplings between its electrical, magnetic and mechanical dynamics, and are satisfied by a large number of practical

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

Passive House Solutions

Energy Technology Data Exchange (ETDEWEB)

Strom, I.; Joosten, L.; Boonstra, C. [DHV Sustainability Consultants, Eindhoiven (Netherlands)

2006-05-15

PEP stands for 'Promotion of European Passive Houses' and is a consortium of European partners, supported by the European Commission, Directorate General for Energy and Transport. In this working paper an overview is given of Passive House solutions. An inventory has been made of Passive House solutions for new build residences applied in each country. Based on this, the most common basic solutions have been identified and described in further detail, including the extent to which solutions are applied in common and best practice and expected barriers for the implementation in each country. An inventory per country is included in the appendix. The analysis of Passive House solutions in partner countries shows high priority with regard to the performance of the thermal envelope, such as high insulation of walls, roofs, floors and windows/ doors, thermal bridge-free construction and air tightness. Due to the required air tightness, special attention must be paid to indoor air quality through proper ventilation. Finally, efficient ((semi-)solar) heating systems for combined space and DHW heating still require a significant amount of attention in most partner countries. Other basic Passive House solutions show a smaller discrepancy with common practice and fewer barriers have been encountered in partner countries. In the next section, the general barriers in partner countries have been inventoried. For each type of barrier a suggested approach has been given. Most frequently encountered barriers in partner countries are: limited know-how; limited contractor skills; and acceptation of Passive Houses in the market. Based on the suggested approaches to overcoming barriers, this means that a great deal of attention must be paid to providing practical information and solutions to building professionals, providing practical training to installers and contractors and communication about the Passive House concept to the market.

Photoresponse and Field Effect Transport Studies in InAsP-InP Core-Shell Nanowires

Science.gov (United States)

Lee, Rochelle; Jo, Min Hyeok; Kim, TaeWan; Kim, Hyo Jin; Kim, Doo Gun; Shin, Jae Cheol

2018-05-01

A ternary InAsyP1-y alloy is suitable for an application to near-infrared (NIR) optical devices as their direct bandgap energy covers the entire NIR band. A nanowire (NW) system allows an epitaxial integration of InAsyP1-y alloy on any type of substrate since the lattice mismatch strain can be relieved through the NW sidewall. Nevertheless, the very large surface to volume ratio feature of the NWs leads to enormous surface states which are susceptible to surface recombination of free carriers. Here, ternary InAs0.75P0.25 NWs are grown with InP passivation layer (i.e., core-shell structure) to minimize the influence of the surface states, thus increasing their optical and electrical properties. A photoresponse study was achieved through the modeled band structure of the grown NWs. The model and experimental results suggest that 5-nm-thick InP shell efficiently passivates the surface states of the InAs0.75P0.25 NWs. The fabricated core-shell photodetectors and field-effect transistors exhibit improved photoresponse and transport properties compared to its counterpart core-only structure.

Photoresponse and Field Effect Transport Studies in InAsP-InP Core-Shell Nanowires

Science.gov (United States)

Lee, Rochelle; Jo, Min Hyeok; Kim, TaeWan; Kim, Hyo Jin; Kim, Doo Gun; Shin, Jae Cheol

2018-03-01

A ternary InAsyP1-y alloy is suitable for an application to near-infrared (NIR) optical devices as their direct bandgap energy covers the entire NIR band. A nanowire (NW) system allows an epitaxial integration of InAsyP1-y alloy on any type of substrate since the lattice mismatch strain can be relieved through the NW sidewall. Nevertheless, the very large surface to volume ratio feature of the NWs leads to enormous surface states which are susceptible to surface recombination of free carriers. Here, ternary InAs0.75P0.25 NWs are grown with InP passivation layer (i.e., core-shell structure) to minimize the influence of the surface states, thus increasing their optical and electrical properties. A photoresponse study was achieved through the modeled band structure of the grown NWs. The model and experimental results suggest that 5-nm-thick InP shell efficiently passivates the surface states of the InAs0.75P0.25 NWs. The fabricated core-shell photodetectors and field-effect transistors exhibit improved photoresponse and transport properties compared to its counterpart core-only structure.

A hybrid damping method for LLCL-filter based grid-tied inverter with a digital filter and an RC parallel passive damper

DEFF Research Database (Denmark)

Wu, Weimin; Lin, Zhe; Sun, Yunjie

2013-01-01

Grid-tied inverters have been widely used to inject the renewable energies into the distributed power generation systems. However, a large variation of the grid impedance challenges the stability of the high-order power filter based grid-tied inverter. Many passive and active damping methods have...... been proposed to overcome this issue. Recently, a composite passive damping method for a high-order power filter based grid-tied inverter with an RC parallel damper and an RL series damper was presented to eliminate this problem, but at the cost of more material and power losses. In this paper...

THE USE OF PASSIVE SOLAR HEATING SYSTEMS AS PART OF THE PASSIVE HOUSE

Directory of Open Access Journals (Sweden)

Bryzgalin Vladislav Viktorovich

2018-05-01

Full Text Available Subject: systems of passive solar heating, which can, without the use of engineering equipment, capture and accumulate the solar heat used for heating buildings. Research objectives: study of the possibility to reach the passive house standard (buildings with near zero energy consumption for heating in climatic conditions of Russia using the systems of passive solar heating in combination with other solutions for reduction of energy costs of building developed in the past. Materials and methods: search and analysis of literature, containing descriptions of various passive solar heating systems, examples of their use in different climatic conditions and the resulting effect obtained from their use; analysis of thermophysical processes occurring in these systems. Results: we revealed the potential of using the solar heating systems in the climatic conditions of parts of the territories of the Russian Federation, identified the possibility of cheaper construction by the passive house standard with the use of these systems. Conclusions: more detailed analysis of thermophysical and other processes that take place in passive solar heating systems is required for creation of their computational models, which will allow us to more accurately predict their effectiveness and seek the most cost-effective design solutions, and include them in the list of means for achieving the passive house standard.

Passive solar construction handbook

Energy Technology Data Exchange (ETDEWEB)

Levy, E.; Evans, D.; Gardstein, C.

1981-08-01

Many of the basic elements of passive solar design are reviewed. The unique design constraints presented in passive homes are introduced and many of the salient issues influencing design decisions are described briefly. Passive solar construction is described for each passive system type: direct gain, thermal storage wall, attached sunspace, thermal storage roof, and convective loop. For each system type, important design and construction issues are discussed and case studies illustrating designed and built examples of the system type are presented. Construction details are given and construction and thermal performance information is given for the materials used in collector components, storage components, and control components. Included are glazing materials, framing systems, caulking and sealants, concrete masonry, concrete, brick, shading, reflectors, and insulators. The Load Collector Ratio method for estimating passive system performance is appended, and other analysis methods are briefly summarized. (LEW)

Modeling illumination performance of plastic optical fiber passive daylighting system

Energy Technology Data Exchange (ETDEWEB)

Sulaiman, F; Ahmad, A [Universiti Teknologi MARA, Shah Alam (Malaysia). Faculty of Electrical Engineering; Ahmed, A Z [Universiti Teknologi MARA, Shah Alam (Malaysia). Bureau of Reseaarch and Consultancy

2006-12-15

of the most direct methods of utilizing solar energy for energy conservation is to bring natural light indoors to light up an area. This paper reports on the investigation of the feasibility to utilize large core optical fibers to convey and distribute solar light passively throughout residential or commercial structures. The focus of this study is on the mathematical modeling of the illumination performance and the light transmission efficiency of solid core end light fiber for optical day lighting systems. The Meatball simulations features the optical fiber transmittance for glass and plastic fibers, illumination performance over lengths of plastic end-lit fiber, spectral transmission, light intensity loss through the large diameter solid core optical fibers as well as the transmission efficiency of the optical fiber itself. It was found that plastic optical fiber has less transmission loss over the distance of the fiber run which clearly shows that the Plastic Optical Fiber should be optimized for emitting visible light. The findings from the analysis on the performance of large diameter optical fibers for day lighting systems seems feasible for energy efficient lighting system in commercial or residential buildings.

Modeling illumination performance of plastic optical fiber passive daylighting system

International Nuclear Information System (INIS)

Sulaiman, F.; Ahmad, A.; Ahmed, A.Z.

2006-01-01

One of the most direct methods of utilizing solar energy for energy conservation is to bring natural light indoors to light up an area. This paper reports on the investigation of the feasibility to utilize large core optical fibers to convey and distribute solar light passively throughout residential or commercial structures. The focus of this study is on the mathematical modeling of the illumination performance and the light transmission efficiency of solid core end light fiber for optical day lighting systems. The Meatball simulations features the optical fiber transmittance for glass and plastic fibers, illumination performance over lengths of plastic end-lit fiber, spectral transmission, light intensity loss through the large diameter solid core optical fibers as well as the transmission efficiency of the optical fiber itself. It was found that plastic optical fiber has less transmission loss over the distance of the fiber run which clearly shows that the Plastic Optical Fiber should be optimized for emitting visible light. The findings from the analysis on the performance of large diameter optical fibers for day lighting systems seems feasible for energy efficient lighting system in commercial or residential buildings

In planta passive sampling devices for assessing subsurface chlorinated solvents.

Science.gov (United States)

Shetty, Mikhil K; Limmer, Matt A; Waltermire, Kendra; Morrison, Glenn C; Burken, Joel G

2014-06-01

Contaminant concentrations in trees have been used to delineate groundwater contaminant plumes (i.e., phytoscreening); however, variability in tree composition hinders accurate measurement of contaminant concentrations in planta, particularly for long-term monitoring. This study investigated in planta passive sampling devices (PSDs), termed solid phase samplers (SPSs) to be used as a surrogate tree core. Characteristics studied for five materials included material-air partitioning coefficients (Kma) for chlorinated solvents, sampler equilibration time and field suitability. The materials investigated were polydimethylsiloxane (PDMS), low-density polyethylene (LDPE), linear low-density polyethylene (LLDPE), polyoxymethylene (POM) and plasticized polyvinyl chloride (PVC). Both PDMS and LLDPE samplers demonstrated high partitioning coefficients and diffusivities and were further tested in greenhouse experiments and field trials. While most of the materials could be used for passive sampling, the PDMS SPSs performed best as an in planta sampler. Such a sampler was able to accurately measure trichloroethylene (TCE) and tetrachloroethylene (PCE) concentrations while simultaneously incorporating simple operation and minimal impact to the surrounding property and environment. Copyright © 2013 Elsevier Ltd. All rights reserved.

Detent Force Reduction of a C-Core Linear Flux-Switching Permanent Magnet Machine with Multiple Additional Teeth

Directory of Open Access Journals (Sweden)

Yi Du

2017-03-01

Full Text Available C-core linear flux-switching permanent magnet (PM machines (LFSPMs are attracting more and more attention due to their advantages of simplicity and robustness of the secondary side, high power density and high torque density, in which both PMs and armature windings are housed in the primary side. The primary salient tooth wound with a concentrated winding consists of C-shaped iron core segments between which PMs are sandwiched and the magnetization directions of these PMs are adjacent and alternant in the horizontal direction. On the other hand, the secondary side is composed of a simple iron core with salient teeth so that it is very suitable for long stroke applications. However, the detent force of the C-core LFSPM machine is relatively high and the magnetic circuit is unbalanced due to the end effect. Thus, a new multiple additional tooth which consists of an active and a traditional passive additional tooth, is employed at each end side of the primary in this paper, so that the asymmetry due to end effect can be depressed and the detent force can be reduced by adjusting the passive additional tooth position. By using the finite element method, the characteristics and performances of the proposed machine are analyzed and verified.

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

Surface Passivation in Empirical Tight Binding

Science.gov (United States)

He, Yu; Tan, Yaohua; Jiang, Zhengping; Povolotskyi, Michael; Klimeck, Gerhard; Kubis, Tillmann

2016-03-01

Empirical Tight Binding (TB) methods are widely used in atomistic device simulations. Existing TB methods to passivate dangling bonds fall into two categories: 1) Method that explicitly includes passivation atoms is limited to passivation with atoms and small molecules only. 2) Method that implicitly incorporates passivation does not distinguish passivation atom types. This work introduces an implicit passivation method that is applicable to any passivation scenario with appropriate parameters. This method is applied to a Si quantum well and a Si ultra-thin body transistor oxidized with SiO2 in several oxidation configurations. Comparison with ab-initio results and experiments verifies the presented method. Oxidation configurations that severely hamper the transistor performance are identified. It is also shown that the commonly used implicit H atom passivation overestimates the transistor performance.

Improved fueling and transport barrier formation with pellet injection from different locations on DIII-D

International Nuclear Information System (INIS)

Baylor, L.R.; Jernigan, T.C.; Gohil, P.

2001-01-01

Pellet injection has been employed on DIII-D from different injection locations to optimize the mass deposition for density profile control and internal transport barrier formation. Transport barriers have been formed deep in the plasma core with central mass deposition from high field side (HFS) injected pellets and in the edge with pellets that trigger L-mode to H-mode transitions. Pellets injected from all locations can trigger the H-mode transition, which depends on the edge density gradient created and not on the radial extent of the pellet deposition. Pellets injected from inside the magnetic axis from the inner wall or vertical port lead to stronger central mass deposition than pellets injected from the low field side (LFS) and thus yield deeper more efficient fueling. (author)

Effects of Core Cavity on a Flow Distribution

Energy Technology Data Exchange (ETDEWEB)

Kwon, Tae-Soon; Kim, Kihwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The axial pressure drop is removed in the free core condition, But the actual core has lots of fuel bundles and mixing vanes to the flow direction. The axial pressure drop induces flow uniformity. In a uniform flow having no shear stress, the cross flow or cross flow mixing decreases. The mixing factor is important in the reactor safety during a Steam Line Break (SLB) or Main Steam Line Break (MSLB) transients. And the effect of core cavity is needed to evaluate the realistic core mixing factor quantification. The multi-dimensional flow mixing phenomena in a core cavity has been studied using a CFD code. The 1/5-scale model was applied for the reactor flow analysis. A single phase water flow conditions were considered for the 4-cold leg and DVI flows. To quantify the mixing intensity, a boron scalar was introduced to the ECC injection water at cold legs and DVI nozzles. The present CFD pre-study was performed to quantify the effects of core structure on the mixing phenomena. The quantified boron mixing scalar in the core simulator model represented the effect of core cavity on the core mixing phenomena. This simulation results also give the information for sensor resolution to measure the boron concentration in the experiments and response time to detect mixing phenomena at the core and reactor vessel.

Passive Heat Exposure Alters Perception and Executive Function

Directory of Open Access Journals (Sweden)

Rachel A. Malcolm

2018-05-01

Full Text Available Findings regarding the influence of passive heat exposure on cognitive function remain equivocal due to a number of methodological issues including variation in the domains of cognition examined. In a randomized crossover design, forty-one male participants completed a battery of cognitive function tests [Visual Search, Stroop, Corsi Blocks and Rapid Visual Information Processing (RVIP tests] prior to and following 1 h of passive rest in either hot (39.6 ± 0.4°C, 50.8 ± 2.3% Rh or moderate (21.2 ± 1.8°C, 41.9 ± 11.4% Rh conditions. Subjective feelings of heat exposure, arousal and feeling were assessed alongside physiological measures including core temperature, skin temperature and heart rate, at baseline and throughout the protocol. Response times were slower in the hot trial on the simple (main effect of trial, P < 0.001 and complex (main effect of trial, P < 0.001 levels of the Stroop test (Hot: 872 ± 198 ms; Moderate: 834 ± 177 ms and the simple level of the visual search test (Hot: 354 ± 54 ms; Moderate: 331 ± 47 ms (main effect of trial, P < 0.001. Participants demonstrated superior accuracy on the simple level of the Visual Search test in the hot trial (Hot: 98.5 ± 3.1%; Moderate: 97.4 ± 3.6% (main effect of trial, P = 0.035. Participants also demonstrated an improvement in accuracy on the complex level of the visual search test following 1 h passive heat exposure (Pre: 96.8 ± 5.9%; Post: 98.1 ± 3.1%, whilst a decrement was seen across the trial in the moderate condition (Pre: 97.7 ± 3.5; Post: 97.0 ± 5.1% (time*trial interaction, P = 0.029. No differences in performance were observed on the RVIP or Corsi Blocks tests (all P > 0.05. Subjective feelings of thermal sensation and felt arousal were higher, feeling was lower in the hot trial, whilst skin temperature, core temperature and heart rate were higher (main effects of trial, all P < 0.001. The findings of the present study suggest that response times for perception

Passive-solar construction handbook

Energy Technology Data Exchange (ETDEWEB)

Levy, E.; Evans, D.; Gardstein, C.

1981-02-01

Many of the basic elements of passive solar design are reviewed. Passive solar construction is covered according to system type, each system type discussion including a general discussion of the important design and construction issues which apply to the particular system and case studies illustrating designed and built examples of the system type. The three basic types of passive solar systems discussed are direct gain, thermal storage wall, and attached sunspace. Thermal performance and construction information is presented for typical materials used in passive solar collector components, storage components, and control components. Appended are an overview of analysis methods and a technique for estimating performance. (LEW)

Passivity and passivity breakdown of 304L stainless steel in hot and concentrated nitric acid

International Nuclear Information System (INIS)

Gillard-Tcharkhtchi, Elsa

2014-01-01

The objective of this study is to characterize the oxidation behavior of 304L stainless steel (SS) in representative conditions of spent nuclear fuel reprocessing, i.e. in hot and concentrated nitric acid. In these conditions the SS electrochemical potential is in the passive domain and its corrosion rate is low. However when the media becomes more aggressive, the potential may be shifted towards the trans-passive domain characterized with a high corrosion rate. Passivity and passivity breakdown in the trans-passive domain are of a major interest for the industry. So as to characterize these phenomenons, this work was undertaken with the following representative conditions: a 304L SS from an industrial sheet was studied, the media was hot and concentrated HNO 3 , long term tests were performed. First, the surface of an immersed 304L SS was characterized with several complementary techniques from the micro to the nanometer scale. Then oxidation kinetics was studied in the passive and in the trans-passive domain. The oxidation behavior was studied thanks to weight loss determination and surface analysis. Finally, oxidation evolution as a function of the potential was studied from the passive to the trans-passive domain. In particular, this allowed us to obtain the anodic curve of 304L SS in hot and concentrated and to define precisely the 304L SS limits of in such conditions. (author) [fr

Emergency core cooling system

International Nuclear Information System (INIS)

Ando, Masaki.

1987-01-01

Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

Passivation of fluorinated activated charcoal

Energy Technology Data Exchange (ETDEWEB)

Del Cul, G.D.; Trowbridge, L.D.; Simmons, D.W.; Williams, D.F.; Toth, L.M.

1997-10-01

The Molten Salt Reactor Experiment (MSRE), at the Oak Ridge National Laboratory has been shut down since 1969 when the fuel salt was drained from the core into two Hastelloy N tanks at the reactor site. In 1995, a multiyear project was launched to remediate the potentially hazardous conditions generated by the movement of fissile material and reactive gases from the storage tanks into the piping system and an auxiliary charcoal bed (ACB). The top 12 in. of the ACB is known by gamma scan and thermal analysis to contain about 2.6 kg U-233. According to the laboratory tests, a few feet of fluorinated charcoal are believed to extend beyond the uranium front. The remainder of the ACB should consist of unreacted charcoal. Fluorinated charcoal, when subjected to rapid heating, can decompose generating gaseous products. Under confined conditions, the sudden exothermic decomposition can produce high temperatures and pressures of near-explosive characteristics. Since it will be necessary to drill and tap the ACB to allow installation of piping and instrumentation for remediation and recovery activities, it is necessary to chemically convert the reactive fluorinated charcoal into a more stable material. Ammonia can be administered to the ACB as a volatile denaturing agent that results in the conversion of the C{sub x}F to carbon and ammonium fluoride, NH{sub 4}F. The charcoal laden with NH{sub 4}F can then be heated without risking any sudden decomposition. The only consequence of heating the treated material will be the volatilization of NH{sub 4}F as a mixture of NH{sub 3} and HF, which would primarily recombine as NH{sub 4}F on surfaces below 200 C. The planned scheme for the ACB denaturing is to flow diluted ammonia gas in steps of increasing NH{sub 3} concentration, 2% to 50%, followed by the injection of pure ammonia. This report summarizes the planned passivation treatment scheme to stabilize the ACB and remove the potential hazards. It also includes basic information

The status of work in the USSR on using inherent self-protection features of fast reactors, of passive and active means of shutdown and decay heat removal system

International Nuclear Information System (INIS)

Buksha, Yu.K.

1991-01-01

Extensive studies on fast reactor safety, aimed to increase intrinsic safety features and introduce passive safety means, are under way in the USSR. Design of the BN-800 reactor core with a close-to-zero sodium void effect of reactivity has been developed, complementary reactivity control means, based on passive principles are being implemented. As a whole, after the Chernobyl accident, the preference is given to the 'passive' full proof methods of safety. This approach may possibly seem excessive and may result in some losses concerning reactor economic characteristics

Electromagnetically controlled measuring device for measuring injection quantities in a diesel injection pump volumetrically. Elektromagnetisch gesteuerte Messvorrichtung zur volumetrischen Messung von Einspritzmengen einer Dieseleinspritzpumpe

Energy Technology Data Exchange (ETDEWEB)

Hoffmann, K H; Mueller, M; Decker, R; Huber, G

1990-11-22

The invention concerns a measuring device for volumetric measurements of injection quantities of a diesel injection pump which injects its contents into a volumetric chamber controlled electromagnetically by a discharge valve and enclosed by a non-impact gas pressure loaded volumetric vessel and effects a retreating movement of the latter. The device is provided with an inductive path controller fitted with a differential pair of coils containing an axially movable ferromagnetic core. The path controller forms a part of a lifter rod connected to the volumetric vessel. It gives an opening signal to the discharge valve after each retreat of the volumetric vessel and a closing signal as soon as a defined height of suspension corresponding to the original position of the volumetric vessel after its return is reached.

Pellet injection into H-mode ITER plasma with the presence of internal transport barriers

Energy Technology Data Exchange (ETDEWEB)

Leekhaphan, P. [Thammasat University, School of Bio-Chemical Engineering and Technology, Sirindhorn International Institute of Technology (Thailand); Onjun, T. [Thammasat University, School of Manufacturing Systems and Mechanical Engineering, Sirindhorn International Institute of Technology (Thailand)

2011-04-15

The impacts of pellet injection into ITER type-1 ELMy H-mode plasma with the presence of internal transport barriers (ITBs) are investigated using self-consistent core-edge simulations of 1.5D BALDUR integrated predictive modeling code. In these simulations, the plasma core transport is predicted using a combination of a semi-empirical Mixed B/gB anomalous transport model, which can self-consistently predict the formation of ITBs, and the NCLASS neoclassical model. For simplicity, it is assumed that toroidal velocity for {omega}{sub E Multiplication-Sign B} calculation is proportional to local ion temperature. In addition, the boundary conditions are predicted using the pedestal temperature model based on magnetic and flow shear stabilization width scaling; while the density of each plasma species, including both hydrogenic and impurity species, at the boundary are assumed to be a large fraction of its line averaged density. For the pellet's behaviors in the hot plasma, the Neutral Gas Shielding (NGS) model by Milora-Foster is used. It was found that the injection of pellet could result in further improvement of fusion performance from that of the formation of ITB. However, the impact of pellet injection is quite complicated. It is also found that the pellets cannot penetrate into a deep core of the plasma. The injection of the pellet results in a formation of density peak in the region close to the plasma edge. The injection of pellet can result in an improved nuclear fusion performance depending on the properties of pellet (i.e., increase up to 5% with a speed of 1 km/s and radius of 2 mm). A sensitivity analysis is carried out to determine the impact of pellet parameters, which are: the pellet radius, the pellet velocity, and the frequency of injection. The increase in the pellet radius and frequency were found to greatly improve the performance and effectiveness of fuelling. However, changing the velocity is observed to exert small impact.

Pellet injection into H-mode ITER plasma with the presence of internal transport barriers

Science.gov (United States)

Leekhaphan, P.; Onjun, T.

2011-04-01

The impacts of pellet injection into ITER type-1 ELMy H-mode plasma with the presence of internal transport barriers (ITBs) are investigated using self-consistent core-edge simulations of 1.5D BALDUR integrated predictive modeling code. In these simulations, the plasma core transport is predicted using a combination of a semi-empirical Mixed B/gB anomalous transport model, which can self-consistently predict the formation of ITBs, and the NCLASS neoclassical model. For simplicity, it is assumed that toroidal velocity for ω E× B calculation is proportional to local ion temperature. In addition, the boundary conditions are predicted using the pedestal temperature model based on magnetic and flow shear stabilization width scaling; while the density of each plasma species, including both hydrogenic and impurity species, at the boundary are assumed to be a large fraction of its line averaged density. For the pellet's behaviors in the hot plasma, the Neutral Gas Shielding (NGS) model by Milora-Foster is used. It was found that the injection of pellet could result in further improvement of fusion performance from that of the formation of ITB. However, the impact of pellet injection is quite complicated. It is also found that the pellets cannot penetrate into a deep core of the plasma. The injection of the pellet results in a formation of density peak in the region close to the plasma edge. The injection of pellet can result in an improved nuclear fusion performance depending on the properties of pellet (i.e., increase up to 5% with a speed of 1 km/s and radius of 2 mm). A sensitivity analysis is carried out to determine the impact of pellet parameters, which are: the pellet radius, the pellet velocity, and the frequency of injection. The increase in the pellet radius and frequency were found to greatly improve the performance and effectiveness of fuelling. However, changing the velocity is observed to exert small impact.

Passive films at the nanoscale

International Nuclear Information System (INIS)

Maurice, Vincent; Marcus, Philippe

2012-01-01

Highlights: ► Nanoscale data on growth, structure and local properties of passive films reviewed. ► Preferential role of defects of passive films on the corrosion resistance emphasized. ► Effect of grain boundaries on local electronic properties shown by new data. ► Use of atomistic modeling to test mechanistic hypotheses illustrated. - Abstract: The nanometer scale chemical and structural aspects of ultrathin oxide passive films providing self-protection against corrosion to metals and alloys in aqueous environments are reviewed. Data on the nucleation and growth of 2D anodic oxide films, details on the atomic structure and nanostructure of 3D passive films, the preferential role of surface step edges in dissolution in the passive state and the preferential role of grain boundaries of the passive films in passivity breakdown are presented. Future perspectives are discussed, and exemplified by new data obtained on the relationship between the nanostructure of oxide passive films and their local electronic properties. Atomistic corrosion modeling by ab initio density functional theory (DFT) is illustrated by the example of interactions of chloride ions with hydroxylated oxide surfaces, including the role of surface step edges. Data obtained on well-defined substrate surfaces with surface analytical techniques are emphasized.

Passivation Of High-Temperature Superconductors

Science.gov (United States)

Vasquez, Richard P.

1991-01-01

Surfaces of high-temperature superconductors passivated with native iodides, sulfides, or sulfates formed by chemical treatments after superconductors grown. Passivating compounds nearly insoluble in and unreactive with water and protect underlying superconductors from effects of moisture. Layers of cuprous iodide and of barium sulfate grown. Other candidate passivating surface films: iodides and sulfides of bismuth, strontium, and thallium. Other proposed techniques for formation of passivating layers include deposition and gas-phase reaction.

NMR-MPar: A Fault-Tolerance Approach for Multi-Core and Many-Core Processors

Directory of Open Access Journals (Sweden)

Vanessa Vargas

2018-03-01

Full Text Available Multi-core and many-core processors are a promising solution to achieve high performance by maintaining a lower power consumption. However, the degree of miniaturization makes them more sensitive to soft-errors. To improve the system reliability, this work proposes a fault-tolerance approach based on redundancy and partitioning principles called N-Modular Redundancy and M-Partitions (NMR-MPar. By combining both principles, this approach allows multi-/many-core processors to perform critical functions in mixed-criticality systems. Benefiting from the capabilities of these devices, NMR-MPar creates different partitions that perform independent functions. For critical functions, it is proposed that N partitions with the same configuration participate of an N-modular redundancy system. In order to validate the approach, a case study is implemented on the KALRAY Multi-Purpose Processing Array (MPPA-256 many-core processor running two parallel benchmark applications. The traveling salesman problem and matrix multiplication applications were selected to test different device’s resources. The effectiveness of NMR-MPar is assessed by software-implemented fault-injection. For evaluation purposes, it is considered that the system is intended to be used in avionics. Results show the improvement of the application reliability by two orders of magnitude when implementing NMR-MPar on the system. Finally, this work opens the possibility to use massive parallelism for dependable applications in embedded systems.

Prediction of the extent of formation damage caused by water injection

Energy Technology Data Exchange (ETDEWEB)

Al-Homadhi, Emad S. [King Saud Univ., Riyadh (Saudi Arabia). Petroleum Engineering Dept.

2013-06-15

As a general practice water is injected along the O/W contact to maintain reservoir pressure during production. Down hole analysis of the injected water shows that, even after surface treatment, it still can contain a considerable amount of solid particles. These particles can bridge formation pores and cause a considerable reduction in the injectivity. To ensure good injectivity over a longer term, the concentration and size of these solids should not exceed certain limits. In this article core flood tests were carried out to simulate high rate injectors. The injected brine contained solid particles in different concentrations and sizes. Particle concentration was between 5 and 20 ppm and the particle mean size was between 2 and 9 {mu}m. The results were presented as damaging ratio versus pore volume injected. Contrarily to previous studies instead of using experimental results in calibrating or evaluating certain theoretical models, the results in this study were directly fitted to produce equations which can predict the extent of damage caused by injected water by knowing the mean size and concentration of the solid particles contained in that water. (orig.)