Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

International Nuclear Information System (INIS)

Ramirez G, C.; Chavez M, C.

2012-10-01

The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

Depressurization test on hot gas duct

International Nuclear Information System (INIS)

Tanihira, Masanori; Kunitomi; Kazuhiko; Inagaki, Yoshiyuki; Miyamoto, Yoshiaki; Sato, Yutaka.

1989-05-01

To study the integrity of internal structures and the characteristics in a hot gas duct under the rapid depressurization accident, depressurization tests have been carried out using a test apparatus installed the hot gas duct with the same size and the same structures as that of the High Temperature Engineering Test Reactor (HTTR). The tests have been performed with three parameters: depressurization rate (0.14-3.08 MPa/s) determined by orifice diameter, area of the open space at the slide joint (11.9-2036 mm 2 ), and initial pressure (1.0-4.0 MPa) filled up in a pressure vessel, by using nitrogen gas and helium gas. The maximum pressure difference applied on the internal structures of the hot gas duct was 2.69 MPa on the liner tube and 0.45 MPa on the separating plate. After all tests were completed, the hot gas duct which was used in the tests was disassembled. Inspection revealed that there were no failure and no deformation on the internal structures such as separating plates, insulation layers, a liner tube and a pressure tube. (author)

Energy penalties associated with the use of a sub-slab depressurization system

International Nuclear Information System (INIS)

Clarkin, M.; Brennan, T.; Osborne, M.C.

1990-01-01

One of the primary radon mitigation techniques used to reduce indoor radon concentrations in houses is a sub-slab depressurization system. In this type of system, a fan removes soil gases containing radon from beneath the floor slab and exhausts the gases to the outdoors by creating a pressure field beneath the slab that is negative relative to the basement air pressure. Because of this negative pressure, indoor conditioned air can be drawn through the floor penetrations and exhausted outdoors. In order to determine the amount of conditioned air that is being lost, a series of experiments utilizing tracer gases were performed in three houses. This paper presents the results of these experiments

Effect of natural circulation on RCS depressurization strategy in PWR NPP

International Nuclear Information System (INIS)

Zhang Kun; Tong Lili; Cao Xuewu

2009-01-01

The natural circulation model of Chinese Qinshan Nuclear Power Plant (NPP) Unit 2 is built using SCDAP/RELAP5 code. Selecting TMLB' accident as the base sequence, this paper analyzes the natural circulation phenomena in high-pressure core melt severe accident. In order to study the effect of natural circulation on RCS depressurization strategy, the accident progressions of RCS depressurization with and without natural circulation are simulated, respectively. According to the results, the natural circulation can delay the initiation of RCS depressurization and the whole accident progression, but it does not evidently influence the results of RCS depressurization. (authors)

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

The effect of the rate of hydrostatic pressure depressurization on cells in culture.

Science.gov (United States)

Tworkoski, Ellen; Glucksberg, Matthew R; Johnson, Mark

2018-01-01

Changes in hydrostatic pressure, at levels as low as 10 mm Hg, have been reported in some studies to alter cell function in vitro; however, other studies have found no detectable changes using similar methodologies. We here investigate the hypothesis that the rate of depressurization, rather than elevated hydrostatic pressure itself, may be responsible for these reported changes. Hydrostatic pressure (100 mm Hg above atmospheric pressure) was applied to bovine aortic endothelial cells (BAECs) and PC12 neuronal cells using pressurized gas for periods ranging from 3 hours to 9 days, and then the system was either slowly (~30 minutes) or rapidly (~5 seconds) depressurized. Cell viability, apoptosis, proliferation, and F-actin distribution were then assayed. Our results did not show significant differences between rapidly and slowly depressurized cells that would explain differences previously reported in the literature. Moreover, we found no detectable effect of elevated hydrostatic pressure (with slow depressurization) on any measured variables. Our results do not confirm the findings of other groups that modest increases in hydrostatic pressure affect cell function, but we are not able to explain their findings.

Vacuum horizontal drainage for depressurization of uranium tailings

International Nuclear Information System (INIS)

Pakalnis, R.; Chedsey, G.; Robertson, A.M.; Follin, S.

1985-01-01

A recent advance in tailings slope depressurization is the application of vacuum assist horizontal drainage. Horizontal drains have been used for several decades to reduce water pressures in slopes in order to improve stability. The benefit from vacuum assist arises from an increased hydraulic gradient caused by induced negative atmospheric pressures. The vacuum assist system has, since its inception in 1982, been successfully employed at two soil and four rock slope projects located in Western Canada. This paper describes the first application of this system in the United States. The technical feasibility of employing vacuum assisted horizontal drains to depressurize a uranium tailings dam near Riverton, Wyoming has been evaluated. Two horizontal drains (300 ft.) were installed and their effect monitored by nine piezometers. The study was conducted over a three-week internal with vacuum being applied for three and four day periods. The drawdowns achieved through vacuum drainage was found to be approximately double that obtained by gravity alone. The volume of water exhausted under vacuum during the seven day interval was approximately double that obtained by gravity alone

Analysis of design and operational effects of filtered containment venting on depressurization and fission product release

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Woon; Seol, Wook-Cheol; Kim, Jisu [Dongguk Univ., Gyeongbuk (Korea, Republic of)

2017-03-15

Effects of design and operational parameters of filtered containment venting system during a specified containment depressurization and relative aero sol release amount are analyzed. The analyses is performed by using the MAAP4 code for the APR1400 reactor. Major results uniquely identified from the analyses can be noted as following: Even though containment depressurization is accelerated as the pipe size increases, the venting system solution is also depleted earlier. Elapsed times to reach lower end pressure of 2 bar are nearly identical regardless of the vent initiation pressure and thus early venting is not much beneficial than late venting. Stroke time of the isolation valves has no effect on the depressurization performance and thus slow opening is beneficial for load reduction from the vent effluent.

A depressurization assistance control based on the posture of a seated patient on a wheelchair.

Science.gov (United States)

Chugo, Daisuke; Fujita, Kazuya; Sakaida, Yuki; Yokota, Sho; Takase, Kunikatsu

2011-01-01

For reducing the risk of pressure sore caused by long period sitting on a wheelchair, we develop a depressurization motion assistance system which is low cost and suitable for practical use. Our developing system consists of a seating cushion which the patient sits on and four air cells which can lift or incline the seating cushion. Each air cell is actuated by small air compressor, which can drive using batteries on the wheelchair respectively, and each compressor has a pressure sensor on its body. In this paper, our key ideas are two topics. One topic is mechanical design for practical use. We realize thin mechanism which enables easy implementation to the general wheelchair. For realizing this thinly design, we develop the tilt mechanism using elasticity of acrylic resin and the controller which uses only pressure sensors for estimating its lifting height and inclination. The other topic is assistance control scheme based on the patient's depressurization operation for increasing a rehabilitation performance. For realizing the proposed control scheme, we analyze the hip depressurization operation by the nursing specialists and use its results for estimating the patient's condition. Using our system, the patient can depressurize by his own will on the general wheelchair easily. The performance of our system is verified by experiments using our prototype. © 2011 IEEE

Mechanical Properties of Porous Titanium Structure Fabricated by Investment Casting with Pressurization/Depressurization System

International Nuclear Information System (INIS)

Kang, San; Lee, Ji-Woon; Hyun, Soong-Keun; Lee, Byong-Pil; Kim, Myoung-Gyun; Kim, Young-Jig

2014-01-01

A porous titanium structure was fabricated by investment casting with a pressurization/depressurization system, and its mechanical properties were studied. A Micro-Vickers hardness profile revealed that hardness gradually increased from the matrix to the metal/mold interface. A compression test was conducted on a single cell of the porous Ti structure. The theoretical and experimental values of yield strength were in good agreement. Such agreement suggested that the reaction layer did not affect the macro-mechanical properties of the porous Ti structure.

Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

2017-05-15

Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

International Nuclear Information System (INIS)

Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

1997-01-01

The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

Study on depressurization measurements and effect in PWR

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

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

Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

International Nuclear Information System (INIS)

Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

2007-01-01

As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

Experimental Investigation on the Behavior of Supercritical CO2 during Reservoir Depressurization.

Science.gov (United States)

Li, Rong; Jiang, Peixue; He, Di; Chen, Xue; Xu, Ruina

2017-08-01

CO 2 sequestration in saline aquifers is a promising way to address climate change. However, the pressure of the sequestration reservoir may decrease in practice, which induces CO 2 exsolution and expansion in the reservoir. In this study, we conducted a core-scale experimental investigation on the depressurization of CO 2 -containing sandstone using NMR equipment. Three different series of experiments were designed to investigate the influence of the depressurization rate and the initial CO2 states on the dynamics of different trapping mechanisms. The pressure range of the depressurization was from 10.5 to 4.0 MPa, which covered the supercritical and gaseous states of the CO 2 (named as CO 2 (sc) and CO 2 (g), respectively). It was found that when the aqueous phase saturated initially, the exsolution behavior strongly depended on the depressurization rate. When the CO 2 and aqueous phase coexisting initially, the expansion of the CO 2 (sc/g) contributed to the incremental CO 2 saturation in the core only when the CO 2 occurred as residually trapped. It indicates that the reservoir depressurization has the possibility to convert the solubility trapping to the residual trapping phase, and/or convert the residual trapping to mobile CO 2 .

RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Lim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [Korea Electric Power Cooperation, Taejon (Korea, Republic of)

1999-12-31

The Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown phenomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifics size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation. 6 refs., 5 figs., 1 tab. (Author)

RCGVS design improvement and depressurization capability tests for Ulchin nuclear power plant units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Sung, Kang Sik; Seong, Ho Je; Jeong, Won Sang; Seo, Jong Tae; Lee, Sang Keun [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Lim, Keun Hyo; Choi, Kwon Sik; Oh, Chul Sung [Korea Electric Power Cooperation, Taejon (Korea, Republic of)

1998-12-31

The Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) has been improved from the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 and 4) based on the evaluation results for depressurization capability tests performed at YGN 3 and 4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown phenomena in order to optimize the orifice size of UCN 3 and 4 RCGVS. Based on these analyses results, the RCGVS orifics size for UCN 3 and 4 has been reduced to 9/32 inch from the 11/32 inch for YGN 3 and 4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3 and 4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation. 6 refs., 5 figs., 1 tab. (Author)

Analysis of Depressurization Performance in Containment of Wolsong NPP Unit 1 through Containment Filtered Venting System

International Nuclear Information System (INIS)

Lee, Sunghan; Kim, Jinhyuck; Suh, Nam Duk; Cho, Songwon

2014-01-01

Containment filtered venting system (CFVS) is designed to open and to close isolation valves passively by an operator. CFVS is operated when the containment pressure exceeds the design pressure (225 kPa(a)) and is closed when the containment pressure decreases below 151 kPa(a). The aim of this study is to analyze the depressurization performance of Wolsong unit 1 through CFVS during SBO. The thermal-hydraulic behavior in containment of Wolsong unit 1 was evaluated using the MELCOR 1.8.6 code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). In addition, in order to evaluate the effects of the CFVS according to the venting area, a sensitivity study depending on different venting area of the CFVS was conducted. Finally, an analysis of the effects of filtering and scrubbing of radioactive material for CFVS is important but not treated in this paper. The SBO accident is chosen to analyze the thermal-hydraulic behavior of Wolsong unit 1. During SBO, the analysis of CFVS affecting on the depressurization of the containment was conducted using MELCOR 1.8.6 code. Also, a sensitivity study was carried out to evaluate the depressurization performance according to the venting area of CFVS. The results show that the containment pressure is considerably decreased and the integrity of the containment could be maintained in case of CFVS operating. Therefore, CFVS has the capacity to keep the containment pressure below the design pressure during SBO. In addition, there are large differences in the containment pressure depending on venting area. We found that the decreasing rate of the pressure in the containment and water level in CFVS depends on the venting area. In the future, a proper requirement for CFVS sizing criteria according to accident scenarios such as LBLOCA, SBLOCA and SGTR, etc. should be evaluated in order to review the licensing for CFVS. Finally, analyses of aerosols, fission product, and radioactive material

Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

International Nuclear Information System (INIS)

Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

1982-01-01

A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

DSMC Simulations of Disturbance Torque to ISS During Airlock Depressurization

Science.gov (United States)

Lumpkin, F. E., III; Stewart, B. S.

2015-01-01

The primary attitude control system on the International Space Station (ISS) is part of the United States On-orbit Segment (USOS) and uses Control Moment Gyroscopes (CMG). The secondary system is part of the Russian On orbit Segment (RSOS) and uses a combination of gyroscopes and thrusters. Historically, events with significant disturbances such as the airlock depressurizations associated with extra-vehicular activity (EVA) have been performed using the RSOS attitude control system. This avoids excessive propulsive "de-saturations" of the CMGs. However, transfer of attitude control is labor intensive and requires significant propellant. Predictions employing NASA's DSMC Analysis Code (DAC) of the disturbance torque to the ISS for depressurization of the Pirs airlock on the RSOS will be presented [1]. These predictions were performed to assess the feasibility of using USOS control during these events. The ISS Pirs airlock is vented using a device known as a "T-vent" as shown in the inset in figure 1. By orienting two equal streams of gas in opposite directions, this device is intended to have no propulsive effect. However, disturbance force and torque to the ISS do occur due to plume impingement. The disturbance torque resulting from the Pirs depressurization during EVAs is estimated by using a loosely coupled CFD/DSMC technique [2]. CFD is used to simulate the flow field in the nozzle and the near field plume. DSMC is used to simulate the remaining flow field using the CFD results to create an in flow boundary to the DSMC simulation. Due to the highly continuum nature of flow field near the T-vent, two loosely coupled DSMC domains are employed. An 88.2 cubic meter inner domain contains the Pirs airlock and the T-vent. Inner domain results are used to create an in flow boundary for an outer domain containing the remaining portions of the ISS. Several orientations of the ISS solar arrays and radiators have been investigated to find cases that result in minimal

Depressurization accidents in a medium-sized high-temperature gas reactor

International Nuclear Information System (INIS)

Ron, S.; Tzoref, J.; Gal, D.

1992-01-01

The amount of fission product release during a core heatup accident in a medium-sized high-temperature gas reactor depends on the size of the inadvertent opening in the primary circuit; this dependence is assessed. The opening triggers a depressurization event that is assumed to be coupled with the failure of the forced circulation in both decay-heat removal systems. The scenario investigated is a beyond-design-base accident. The DSNP modular simulation code is used. This paper reports that a two-dimensional model is developed to simulate the HTR-500 design. The study shows that the depressurization process does not contribute significantly to the sweeping out (from the primary circuit) of fission products released from the fuel during the core heatup. There is also no significant variation in the results when the opening size is >33 cm 2 , and only a slight sensitivity is found when the rupture size is between 3.3 and 33 cm 2 . The fission product release decreases considerably in the range from 1 to 3.3 cm 2 . The small-sized rupture is of major significance, as the failure of the relief valves to reclose increases the frequency of the event

Development and evaluation of a new depressurization spillage test for residential gas-fired combustion appliances : final report

International Nuclear Information System (INIS)

Edwards, P.

2005-07-01

This paper presented a newly developed combustion depressurization spillage test for residential combustion appliances. The test uses carbon dioxide (CO 2 ) that is produced in the fuel combustion process as a tracer gas. The test accurately measures the amount of combustion spillage from residential combustion appliances and their venting systems when they operate at certain levels of depressurization. Seven commonly used gas-fired appliances were used to evaluate the new test as well as the appliances. These included 2 power-vented storage-tank water heaters, 1 mid-efficiency furnace, 2 high-efficiency condensing furnaces, and 2 direct-vent gas fireplaces. Tests were performed for each unit with the test room initially depressurized by 50 Pa compared with the pressure outside the room. If the combustion spillage exceeded 2 per cent, the test was repeated with the room depressurized by 20 Pa, and then by 5 Pa. Each appliance was operated for 5 minutes of burner operation during which time the burner fuel consumption, the concentration of CO 2 and the exhaust fan flow rate were monitored. Measurements were taken for 2 minutes following burner shut off. The amount of CO 2 that was released into the test room from the appliance and its venting system was determined from the measurements and then compared with the amount of CO 2 that would be produced by combustion of the fuel that was consumed during the test. The ratio of the 2 provided a direct measure of the combustion spillage of the appliance and its venting system. The study revealed that 3 products had undetectable levels of combustion spillage, 3 products had low, but measurable combustion spillage, and 1 product had significant combustion spillage. refs., tabs., figs

Rapid depressurization of a compressible fluid

International Nuclear Information System (INIS)

Dang, M.; Dupont, J.F.; Weber, H.

1978-08-01

The rapid depressurization of a plenum is a situation frequently encountered in the dynamical analysis of nuclear gas cycles of the HHT type. Various methods of numerical analyses for a 1-dimensional flow model are examined: finite difference method; control volume method; method of characteristics. Based on the shallow water analogy to compressible flow, the numerical results are compared with those from a water table set up to simulate a standard problem. (Auth.)

BLOW.MOD2: program for a vessel depressurization calculation with the contribution of structures

International Nuclear Information System (INIS)

Doval, A.

1990-01-01

The BLOW.MOD2 program developed to calculate pressure vessels' depressurization is presented, considering heat contribution of the structures. The results are opposite to those obtained from other more complex numerical models, being the comparison extremely satisfactory. BLOW.MOD2 is a software of the 'Systems Sub-Branch', INVAP S.E. (Author) [es

ACTIVE SOIL DEPRESSURIZATION (ASD) DEMONSTRATION IN A LARGE BUILDING

Science.gov (United States)

The report gives results of an evaluation of the feasibility of implementing radon resistant construction techniques -- especially active soil depressurization (ASD) -- in new large buildings in Florida. Indoor radon concentrations and radon entry were monitored in a finished bui...

Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

Energy Technology Data Exchange (ETDEWEB)

Mueftueoglu, A.K.; Feltus, M.A. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and other signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.

COMEDIE BD1 experiment: Fission product behaviour during depressurization transients

International Nuclear Information System (INIS)

Gillet, R.; Brenet, D.; Hanson, D.L.; Kimball, O.F.

1996-01-01

An experimental program in the CEA COMEDIE loop has been carried out to obtain integral test data to validate the methods and transport models used to predict fission product release from the core and plate-out in the primary coolant circuit of the Modular High Temperature Gas Cooled Reactor (MHTGR) during normal operation and liftoff, and during rapid depressurization transients. The loop consists of an in-pile section with the fuel element, deposition section (heat exchanger), filters for collecting condensible Fission Productions (FP) during depressurization tests and an out-of-pile section devoted to chemical composition control of the gas and on-line analysis of gaseous FP. After steady state irradiation, the loop was subjected to a series of in-situ blowdowns at shear ratios (ratio of the wall shear stress during blowdown to that during steady state operation) ranging from 0.7 to 5.6. The results regarding the FP profiles on the plate-out section, before and after blowdowns are given. It appears that: the plate-out profiles depend on the FP chemistry; the depressurization phases have led to significant desorption of I 131, but on the contrary, they have almost no effect for the other FP such as Ag 110m, Cs 134, Cs 137 and Te 132. (author). 1 ref., 15 figs

Maximum Recoverable Gas from Hydrate Bearing Sediments by Depressurization

KAUST Repository

Terzariol, Marco

2017-11-13

The estimation of gas production rates from hydrate bearing sediments requires complex numerical simulations. This manuscript presents a set of simple and robust analytical solutions to estimate the maximum depressurization-driven recoverable gas. These limiting-equilibrium solutions are established when the dissociation front reaches steady state conditions and ceases to expand further. Analytical solutions show the relevance of (1) relative permeabilities between the hydrate free sediment, the hydrate bearing sediment, and the aquitard layers, and (2) the extent of depressurization in terms of the fluid pressures at the well, at the phase boundary, and in the far field. Close form solutions for the size of the produced zone allow for expeditious financial analyses; results highlight the need for innovative production strategies in order to make hydrate accumulations an economically-viable energy resource. Horizontal directional drilling and multi-wellpoint seafloor dewatering installations may lead to advantageous production strategies in shallow seafloor reservoirs.

Monitoring system of depressurization valves of migrated gas in annular space of flexible risers

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Luiz A.; Santos, Joilson M.; Carvalho, Antonio L.; Loureiro, Patricia [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

2005-07-01

PETROBRAS Research and Development Center - CENPES developed an automatic system for monitoring pressure of annular space due to permeation of gas in flexible risers to inspect continuously integrity of such lines. To help maintaining physical integrity of flexible risers, two PSV's are installed to end fittings on top of riser, so that operation of any valve grants the maximum admissible gas pressure within the riser annular space, as overpressure might cause damages to external polymeric layer of flexible riser. Due to the fact that there is no mechanism allowing operation to verify correct PSV performance and frequency of valve's closings and openings, we felt to be necessary the development and implement an automatic instrumented system, integrated to platform's automation and control infrastructure. The objective of this instrumentation is to monitor and register pressure of annular space in flexible riser, as well as XV's depressurization frequency. Having such information registered and monitored, can infer some riser structural conditions, anticipating repairs and preventive maintenance. In this paper we present developed system details including instruments required, application, operation of associated screens that are used in the ECOS, with events, alarms and industrial automation services required (Application development and system integration). (author)

Depressurization accident analyses for the Fort St. Vrain Reactor

International Nuclear Information System (INIS)

Paul, D.D.

1976-01-01

Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cooling system. Results are presented for the Pelton circulators operating at 10,550, 8800, and 7000 rpm. Maximum temperatures of selected components are plotted as a function of time during the transient. None of the components studied exceeded the temperature at which failure or damage may occur. However, there must be sufficient mixing of the outlet gas in the lower plenum to insure the integrity of the steel liners of the steam generator inlet ducts

Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

International Nuclear Information System (INIS)

Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

2005-01-01

Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

Pore network modelling of heavy oil depressurization : a parametric study of factors affecting critical gas saturation and three-phase relative permeabilities

Energy Technology Data Exchange (ETDEWEB)

Bondino, I.; McDougall, S.D. [Heriot-Watt Univ., Edinburgh, Scotland (United Kingdom); Hamon, G. [TotalFina Elf Exploration and Production (France)

2002-07-01

A review of how the bubble nucleation process affects the efficiency of heavy oil recovery was presented along with a discussion regarding a pore-scale simulator technique to depressurize heavy oil systems. A light oil depressurization simulation is also presented in which a straightforward instantaneous nucleation (IN) model and a more intricate progressive nucleation (PN) model have been used. Simulation results are compared to those derived from the heavy oil systems. The nucleation of bubbles, their growth by solute diffusion and expansion, plus the final stages of coalescence migration and production are the main steps in the depressurization process which were accounted for in a 3-phase simulator. The model can also determine the impact of bubble density and gas-oil diffusion coefficient on critical gas saturation and 3-phase relative permeability. The difference in results for light and heavy oils was also highlighted. In the first scenario, the evolution of gas was characterized by embryonic bubbles that are quickly and randomly nucleated once bubble-point pressure is reached. A stochastic algorithm was developed for PN from experimental observations. IN and PN observations were not necessarily contradictory. It was determined that the high interfacial tension of heavy oils leads to a more compact, capillary-dominated pattern of gas evolution compared to light oils, resulting in improved recoveries for heavy oil systems. 23 refs., 6 tabs., 23 figs.

Emergency core cooling system

International Nuclear Information System (INIS)

Abe, Nobuaki.

1993-01-01

A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients.

Science.gov (United States)

Jun, Jonathan C; Unnikrishnan, Dileep; Schneider, Hartmut; Kirkness, Jason; Schwartz, Alan R; Smith, Philip L; Polotsky, Vsevolod Y

2016-01-01

Obstructive Sleep Apnea (OSA) describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP) is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep. Nine obese (BMI = 40.4 ± 3.5) subjects with severe OSA (AHI = 88.9 ± 6.8) adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20) or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods. 8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (p<0.05). The average 3% ODI during CPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (p<0.05). CPAP depressurization also induced more awake (p<0.05) and stage N1 (p<0.01) sleep, and less stage REM (p<0.05) with a trend towards decreased stage N3 (p = 0.064). Acute intermittent depressurization of CPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP.

Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients.

Directory of Open Access Journals (Sweden)

Jonathan C Jun

Full Text Available Obstructive Sleep Apnea (OSA describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep.Nine obese (BMI = 40.4 ± 3.5 subjects with severe OSA (AHI = 88.9 ± 6.8 adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20 or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods.8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (p<0.05. The average 3% ODI during CPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (p<0.05. CPAP depressurization also induced more awake (p<0.05 and stage N1 (p<0.01 sleep, and less stage REM (p<0.05 with a trend towards decreased stage N3 (p = 0.064.Acute intermittent depressurization of CPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP.

Data report of ROSA/LSTF experiment SB-CL-32. 1% cold leg break LOCA with SG depressurization and no gas inflow

International Nuclear Information System (INIS)

Takeda, Takeshi

2014-11-01

An experiment SB-CL-32 was conducted on May 28, 1996 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-CL-32 simulated a 1% cold leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and no inflow of non-condensable gas from accumulator (ACC) tanks of emergency core cooling system. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after the break. After the initiation of AM action, auxiliary feedwater injection into the SG secondary-side was started with some delay. After the onset of AM action, the primary pressure decreased following the SG secondary-side pressure. Core uncovery by core boil-off started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first loop seal clearing (LSC). The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery by core boil-off took place before second LSC induced by steam condensation on ACC coolant injected into cold legs following the primary depressurization. The core liquid level recovered rapidly after the second LSC. The observed maximum fuel rod surface temperature was 772 K. The experiment was terminated when the continuous core cooling was confirmed because of the coolant injection by low pressure injection system after the isolation of ACC system. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal-hydraulic phenomena. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment SB-CL-32. (author)

An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2006-03-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

Geomechanical response of permafrost-associated hydrate deposits to depressurization-induced gas production

Science.gov (United States)

Rutqvist, J.; Moridis, G.J.; Grover, T.; Collett, T.

2009-01-01

In this simulation study, we analyzed the geomechanical response during depressurization production from two known hydrate-bearing permafrost deposits: the Mallik (Northwest Territories, Canada) deposit and Mount Elbert (Alaska, USA) deposit. Gas was produced from these deposits at constant pressure using horizontal wells placed at the top of a hydrate layer (HL), located at a depth of about 900??m at the Mallik site and 600??m at the Mount Elbert site. The simulation results show that general thermodynamic and geomechanical responses are similar for the two sites, but with substantially higher production and more intensive geomechanical responses at the deeper Mallik deposit. The depressurization-induced dissociation begins at the well bore and then spreads laterally, mainly along the top of the HL. The depressurization results in an increased shear stress within the body of the receding hydrate and causes a vertical compaction of the reservoir. However, its effects are partially mitigated by the relatively stiff permafrost overburden, and compaction of the HL is limited to less than 0.4%. The increased shear stress may lead to shear failure in the hydrate-free zone bounded by the HL overburden and the downward-receding upper dissociation interface. This zone undergoes complete hydrate dissociation, and the cohesive strength of the sediment is low. We determined that the likelihood of shear failure depends on the initial stress state as well as on the geomechanical properties of the reservoir. The Poisson's ratio of the hydrate-bearing formation is a particularly important parameter that determines whether the evolution of the reservoir stresses will increase or decrease the likelihood of shear failure.

Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

Energy Technology Data Exchange (ETDEWEB)

Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2012-10-15

The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

Experimental Investigation Evaporation of Liquid Mixture Droplets during Depressurization into Air Stream

Science.gov (United States)

Liu, L.; Bi, Q. C.; Terekhov, Victor I.; Shishkin, Nikolay E.

2010-03-01

The objective of this study is to develop experimental method to study the evaporation process of liquid mixture droplets during depressurization and into air stream. During the experiment, a droplet was suspended on a thermocouple; an infrared thermal imager was used to measure the droplet surface temperature transition. Saltwater droplets were used to investigate the evaporation process during depressurization, and volatile liquid mixtures of ethanol, methanol and acetone in water were applied to experimentally research the evaporation into air stream. According to the results, the composition and concentration has a complex influence on the evaporation rate and the temperature transition. With an increase in the share of more volatile component, the evaporation rate increases. While, a higher salt concentration in water results in a lower evaporation rate. The shape variation of saltwater droplet also depends on the mass concentration in solution, whether it is higher or lower than the eutectic point (22.4%). The results provide important insight into the complex heat and mass transfer of liquid mixture during evaporation.

Depressurization as a means of leak checking large vacuum vessels

International Nuclear Information System (INIS)

Callis, R.W.; Langhorn, A.; Petersen, P.I.; Ward, C.; Wesley, J.

1985-01-01

A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, the vessel need only be depressurized 2 psi, thus allowing personnel inside to perform the sniffing operation. Second, the sniffing probe used (Leybold--Heraus ''Quick Test'') could detect a change in helium concentration as small as 100 ppb, which allows for faster scanning of the vessel inferior. Use of this technique to find an elusive 10 -3 Torrxl/s leak in the Doublet III tokamak vacuum vessel will be presented

Dynamic solution of vessel depressuring; Simulacao dinamica de despressurizacao em vasos

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, Marco Tulio; Silva Netto, Rafael [Chemtech Servicos de Engenharia e Software Ltda., Rio de Janeiro, RJ (Brazil); Aires, Joyce Stone S. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil). Centro de Pesquisas (CENPES)

2004-07-01

Vessel depressuring is an important phenomenon on chemical and petrochemical processes, specially those related to oil and gas exploration, production and processing. The correct modeling of this phenomenon and prediction of the temperatures, mass and thermal rates involved is essential to the adequate design of the equipment, materials specification and safety standards assurance. To fulfill these requirements, we developed a method to simulate the phenomenon of depressuring. In this approach, the dynamic process is discretized along time, being calculated at each interval the conditions inside the vessel, material flow and heat exchanges with the environment, through mass and energy balances and thermodynamic equilibrium calculation. The method can include different models to calculate heat exchanges and flow through the relief valve, and new methods can be incorporated if necessary. The efficiency of the method was verified by comparing its results with the ones obtained by market-leaders process simulators, when it was proved the robustness of the method and precision of the results. To increase its flexibility of use, the method was incorporated to the PETROBRAS Process Simulator - PETROX (PETROBRAS, 2004), in a development made by Chemtech under a contract with PETROBRAS and already used in large scale by PETROBRAS to simulate its process units. (author)

Effect of Acute Intermittent CPAP Depressurization during Sleep in Obese Patients

Science.gov (United States)

Jun, Jonathan C.; Unnikrishnan, Dileep; Schneider, Hartmut; Kirkness, Jason; Schwartz, Alan R.; Smith, Philip L.; Polotsky, Vsevolod Y.

2016-01-01

Background Obstructive Sleep Apnea (OSA) describes intermittent collapse of the airway during sleep, for which continuous positive airway pressure (CPAP) is often prescribed for treatment. Prior studies suggest that discontinuation of CPAP leads to a gradual, rather than immediate return of baseline severity of OSA. The objective of this study was to determine the extent of OSA recurrence during short intervals of CPAP depressurization during sleep. Methods Nine obese (BMI = 40.4 ± 3.5) subjects with severe OSA (AHI = 88.9 ± 6.8) adherent to CPAP were studied during one night in the sleep laboratory. Nasal CPAP was delivered at therapeutic (11.1 ± 0.6 cm H20) or atmospheric pressure, in alternating fashion for 1-hour periods during the night. We compared sleep architecture and metrics of OSA during CPAP-on and CPAP-off periods. Results 8/9 subjects tolerated CPAP withdrawal. The average AHI during CPAP-on and CPAP-off periods was 3.6 ± 0.6 and 15.8 ± 3.6 respectively (pCPAP-on and CPAP-off was 4.7 ± 2 and 20.4 ± 4.7 respectively (pCPAP depressurization also induced more awake (pCPAP during sleep led to deterioration of sleep architecture but only partial re-emergence of OSA. These observations suggest carryover effects of CPAP. PMID:26731735

Experimental and theoretical investigation on the depressurization of a vessel with internals

International Nuclear Information System (INIS)

Vigni, P.; Oriolo, F.; Rosa, U.

1978-01-01

This paper is about some blow-down experiments performed at the Scalbatraio Center of the University of Pisa. The blow-down tests have been made to investigate the depressurization of a vessel with internal structures, reproducing the geometry of a BWR. The experimental data have been compared with calculations performed by the RELAP program, in order to evaluate the scaling effects related to their application to large scale units. (author)

Reactor containment depressurization and filtration equipment for use in the case of a serious accident

International Nuclear Information System (INIS)

L'Homme, A.

1987-06-01

A study was carried out under the aegis of the OECD into filtered vented containment systems which permit depressurization of the containment and filtration of the effluents released to the environment, in the event of a major accident with a pressurized water reactor (PWR) (or BWR or CANDU type reactors) involving core meltdown, with a view to minimizing the consequences. This paper describes the various systems examined which could possibly be used for this purpose. These comprised the French robust sand filtration system, the Swedish FILTRA system, the vacuum containment and discharge and emergency filtration system used by the CANDU plants of the Ontario-Hydro electricity company in Canada and the BWR pressure-suppression pounds. The positions of the various national authorities regarding incorporation of such systems into nuclear power plants, the design and technical principles underlying the systems, the procedures and criteria for their use and their advantages and disadvantages are examined [fr

Radon remediation of a two-storey UK dwelling by active sub-slab depressurization: observations on hourly Radon concentration variations

International Nuclear Information System (INIS)

Denman, A.R.

2008-01-01

Radon concentration levels in a two-storey detached single-family dwelling in Northamptonshire, UK, were monitored at hourly intervals throughout a 5-week period during which sub-slab depressurization remediation measures, including an active sump system, were installed. Remediation of the property was accomplished successfully, with the mean radon levels upstairs and downstairs greatly reduced and the prominent diurnal variability in radon levels present prior to remediation almost completely removed. Following remediation, upstairs and downstairs radon concentrations were 32% and 16% of their pre-remediation values respectively. The mean downstairs radon concentration was lower than that upstairs, with pre-and post-remediation values of the upstairs/downstairs concentration ratio, R U/D , of 0.93 and 1.76 respectively. Cross-correlation between upstairs and downstairs radon concentration time-series indicates a time-lag of the order of 1 hour or less, suggesting that diffusion of soil-derived radon from downstairs to upstairs either occurs within that time frame or forms a relatively insignificant contribution to the upstairs radon level. Cross-correlation between radon concentration time-series and the corresponding time-series for local atmospheric parameters demonstrated correlation between radon concentrations and internal/external pressure-difference prior to remediation. This correlation disappears following remediation, confirming the effectiveness of the remediation procedure in mitigating radon ingress from the ground via the stack-effect. Overall, these observations provide further evidence that radon emanation from building materials makes a not insignificant contribution to radon concentration levels within the building. Furthermore, since this component remains essentially unaffected by sub-slab depressurization, its proportional contribution to the total radon levels in the home increases following remediation, leading to the conclusion that where

Hydro-geomechanical behaviour of gas-hydrate bearing soils during gas production through depressurization and CO2 injection

Science.gov (United States)

Deusner, C.; Gupta, S.; Kossel, E.; Bigalke, N.; Haeckel, M.

2015-12-01

Results from recent field trials suggest that natural gas could be produced from marine gas hydrate reservoirs at compatible yields and rates. It appears, from a current perspective, that gas production would essentially be based on depressurization and, when facing suitable conditions, be assisted by local thermal stimulation or gas hydrate conversion after injection of CO2-rich fluids. Both field trials, onshore in the Alaska permafrost and in the Nankai Trough offshore Japan, were accompanied by different technical issues, the most striking problems resulting from un-predicted geomechanical behaviour, sediment destabilization and catastrophic sand production. So far, there is a lack of experimental data which could help to understand relevant mechanisms and triggers for potential soil failure in gas hydrate production, to guide model development for simulation of soil behaviour in large-scale production, and to identify processes which drive or, further, mitigate sand production. We use high-pressure flow-through systems in combination with different online and in situ monitoring tools (e.g. Raman microscopy, MRI) to simulate relevant gas hydrate production scenarios. Key components for soil mechanical studies are triaxial systems with ERT (Electric resistivity tomography) and high-resolution local strain analysis. Sand production control and management is studied in a novel hollow-cylinder-type triaxial setup with a miniaturized borehole which allows fluid and particle transport at different fluid injection and flow conditions. Further, the development of a large-scale high-pressure flow-through triaxial test system equipped with μ-CT is ongoing. We will present results from high-pressure flow-through experiments on gas production through depressurization and injection of CO2-rich fluids. Experimental data are used to develop and parametrize numerical models which can simulate coupled process dynamics during gas-hydrate formation and gas production.

Mathematical simulation of the drying of suspensions and colloidal solutions by their depressurization

Science.gov (United States)

Lashkov, V. A.; Levashko, E. I.; Safin, R. G.

2006-05-01

The heat and mass transfer in the process of drying of high-humidity materials by their depressurization has been investigated. The results of experimental investigation and mathematical simulation of the indicated process are presented. They allow one to determine the regularities of this process and predict the quality of the finished product. A technological scheme and an engineering procedure for calculating the drying of the liquid base of a soap are presented.

Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

Science.gov (United States)

Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

2017-10-01

The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

System and method for slurry handling

Science.gov (United States)

Steele, Raymond Douglas; Oppenheim, Judith Pauline

2015-12-29

A system includes a slurry depressurizing system that includes a liquid expansion system configured to continuously receive a slurry at a first pressure and continuously discharge the slurry at a second pressure. For example, the slurry depressurizing system may include an expansion turbine to expand the slurry from the first pressure to the second pressure.

Analytical method and result of radiation exposure for depressurization accident of HTTR

International Nuclear Information System (INIS)

Sawa, K.; Shiozawa, S.; Mikami, H.

1990-01-01

The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR). Since the HTTR has some characteristics different from LWRs, analytical method of radiation exposure in accidents provided for LWRs can not be applied directly. This paper describes the analytical method of radiation exposure developed by JAERI for the depressurization accident, which is the severest accident in respect to radiation exposure among the design basis accidents of the HTTR. The result is also described in this paper

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Corletti, M.M.; Schulz, T.L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

Physical modelling of LNG rollover in a depressurized container filled with water

Science.gov (United States)

Maksim, Dadonau; Denissenko, Petr; Hubert, Antoine; Dembele, Siaka; Wen, Jennifer

2015-11-01

Stable density stratification of multi-component Liquefied Natural Gas causes it to form distinct layers, with upper layer having a higher fraction of the lighter components. Heat flux through the walls and base of the container results in buoyancy-driven convection accompanied by heat and mass transfer between the layers. The equilibration of densities of the top and bottom layers, normally caused by the preferential evaporation of Nitrogen, may induce an imbalance in the system and trigger a rapid mixing process, so-called rollover. Numerical simulation of the rollover is complicated and codes require validation. Physical modelling of the phenomenon has been performed in a water-filled depressurized vessel. Reducing gas pressure in the container to levels comparable to the hydrostatic pressure in the water column allows modelling of tens of meters industrial reservoirs using a 20 cm laboratory setup. Additionally, it allows to model superheating of the base fluid layer at temperatures close the room temperature. Flow visualizations and parametric studies are presented. Results are related to outcomes of numerical modelling.

AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-10-15

Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

International Nuclear Information System (INIS)

Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

2005-01-01

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Schulz, T.L.; Corletti, M.M.

1994-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

Evaluation of the need for a rapid depressurization capability for Combustion Engineering plants

International Nuclear Information System (INIS)

Marsh, L.; Liang, C.

1984-12-01

This report documents the NRC staff evaluation of the need for providing a rapid primary system depressurization capability, in particular by using a power-operated relief valve(s) (PORVs), in the current 3410-MWt and 3800-MWt classes of plants designed by Combustion Engineering (CE). The staff reviewed the responses of licensees, applicants, and vendors to staff questions, supplemented by independent analyses by the staff and its contractors. The staff review led to the conclusion that, on the basis of risk reduction and cost/benefit considerations, no overwhelming benefit would result from requiring the installation of PORVs in CE plants that currently do not have them. However, when other unquantifiable considerations regarding the potential benefits of a PORV are factored into the evaluation, it appears that more substantial benefits could be realized. Given the more comprehensive studies currently under way to resolve the generic unresolved safety issue, USI A-45, Decay Heat Removal Reliability, the staff concludes that the decision regarding PORVs for these CE plants should be deferred and incorporated into the technical resolution of USI A-45

A simple method for environmental cell depressurization for use with an electron microscope.

Science.gov (United States)

Ogawa, Naoki; Mizokawa, Ryo; Saito, Minoru; Ishikawa, Akira

2017-12-01

With the aid of the environmental cell (EC) in electron microscopy, hydrated specimens have been observed at high resolutions that optical microscopy cannot attain. Due to the ultra-high vacuum conditions of the inner column of the electron microscope, the EC requires sealing films that are sufficiently thin to allow electron transmission and that are sufficiently tough to withstand the pressure difference between the inside and outside of the EC. However, most hydrated specimens can be observed at low vacuum because the saturated vapor pressure of water is known to be 0.02 atm at room temperature. These concepts have been used in the differential pumping system, but it is complicated and relatively expensive. In this work, we propose a simple method for depressurization of the EC using a 'balloon structure' and demonstrate the theoretical benefits and practical improvement for specimen observations in low-vacuum conditions. © The Author 2017. Published by Oxford University Press on behalf of The Japanese Society of Microscopy. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Nuclear reactor steam depressurization valve

International Nuclear Information System (INIS)

Moore, G.L.

1991-01-01

This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2005-06-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

Droplet Impact on a Heated Surface under a Depressurized Environment

Science.gov (United States)

Hatakenaka, Ryuta; Tagawa, Yoshiyuki

2016-11-01

Behavior of a water droplet of the diameter 1-3mm impacting on a heated surface under depressurized environment (100kPa -1kPa) has been studied. A syringe pump for droplet generation and a heated plate are set into a transparent acrylic vacuum chamber. The internal pressure of the chamber is automatically controlled at a target pressure with a rotary pump, a pressure transducer, and an electrical valve. A silicon wafer of the thickness 0.28 mm is mounted on the heater plate, whose temperature is directly measured by attaching a thermocouple on the backside. The droplet behavior is captured using a high-speed camera in a direction perpendicular to droplet velocity. Some unique behaviors of droplet are observed by decreasing the environmental pressure, which are considered to be due to two basic elements: Enhancement of evaporation due to the lowered saturation temperature, and shortage of pneumatic spring effect between the droplet and heated wall due to the lowered pressure of the air.

Source Test Report for the 205 Delayed Coking Unit Drum 205-1201 and Drum 205-1202 Depressurization Vents (Marathon Petroleum Company LLC)

Science.gov (United States)

The 2010 Source Test was performed during the atmospheric depressurization step of the delayed coking process prior to the removal of petroleum coke from the coke drum. The 205 DCU was operated under a variety of conditions during the 2010 Source Test.

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

Two-phase mixture level swell and liquid entrainment/off-take in a vessel during rapid depressurization

International Nuclear Information System (INIS)

Kim, Chang Hyun

2004-02-01

An experimental study has been performed to analyze the two-phase mixture level swell and the liquid entrainment/off-take through the break in a vessel, which are important phenomena to determine the bleed capacity of the Safety Depressurization System (SDS) of Korea Advanced Power Reactor 1400 (APR1400). Three separate experiments are performed in this study: (a) the depressurization and two-phase mixture level swell experiment: (b) the two-phase mixture level measurement experiment: (c) the liquid entrainment and off-take experiment. A series of experiments has been performed using a scaled pressurized vessel in various depressurization conditions to analyze the two-phase mixture level swell and the liquid entrainment/off-take phenomena from the two-phase mixture surface in the first experiment. The test parameters are the initial pressure (10 - 38.75bars), the initial water level (43.7% - 80.0% of full height), the orifice inner diameter (10mm, 17.5mm, and 20mm). The liquid off-take takes place in certain experimental conditions. The measured parameters in the present experiments are axial void fraction distributions, pressures, temperatures in the test vessel, and the mixture density and mass flowrate through the discharge pipe. An assessment of RELAP5/MOD3 code with the present experimental data has been performed. With appropriate nodalization and time step, RELAP5/MOD3 showed reasonable agreement with the present experimental data for the gradual depressurization without liquid off-take. In the case that the off-take takes place, however, RELAP5/MOD3 under-predicts the amount of liquid entrainment/off-take during depressurization. In the second experiment, an assessment of an ultrasonic sensor and a two-wire type capacitance probe for the two-phase mixture level measurement has been performed under the same experimental conditions to adopt an appropriate measurement method for the two-phase mixture level swell and to investigate pool void fraction by the

DESIGN AND TESTING OF SUB-SLAB DEPRESSURIZATION FOR RADON MITIGATION IN NORTH FLORIDA HOUSES - PART I. PERFORMANCE AND DURABILITY - VOLUME 1. TECHNICAL REPORT

Science.gov (United States)

The report gives results of a demonstration/research project to evaluate sub-slab depressurization (SSD) techniques for radon mitigation in North Florida where the housing stock is primarily slab-on-grade and the sub-slab medium typically consists of native soil and sand. Objecti...

DESIGN AND TESTING OF SUB-SLAB DEPRESSURIZATION FOR RADON MITIGATION IN NORTH FLORIDA HOUSES - PART I. PERFORMANCE AND DURABILITY - VOLUME 2. DATA APPENDICES

Science.gov (United States)

The report gives results of a demonstration/research project to evaluate sub-slab depressurization (SSD) techniques for radon mitigation in North Florida where the housing stock is primarily slab-on-grade and the sub-slab medium typically consists of native soil and sand. Objecti...

A study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention. ROSA-V/LSTF test SB-PV-05

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2006-11-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system. (author)

The mechanisms of transitions from natural convection and nucleate boiling to nucleate boiling or film boiling caused by rapid depressurization in highly subcooled water

International Nuclear Information System (INIS)

Sakurai, Akira; Shiotsu, Masahiro; Hata, Koichi; Fukuda, Katsuya

1999-01-01

The mechanisms of transient boiling process including the transitions to nucleate boiling or film boiling from initial heat fluxes, q in , in natural convection and nucleate boiling regimes caused by exponentially decreasing system pressure with various decreasing periods, τ p on a horizontal cylinder in a pool of highly subcooled water were clarified. The transient boiling processes with different characteristics were divided into three groups for low and intermediate q in in natural convection regime, and for high q in in nucleate boiling regime. The transitions at maximum heat fluxes from low q in in natural convection regime to stable nucleate boiling regime occurred independently of the τ p values. The transitions from intermediate and high q in values in natural convection and nucleate boiling to stable film boiling occurred for short τ p values, although those to stable nucleate boiling occurred for tong τ p values. The CHF and corresponding surface superheat values at which the transition to film boiling occurred were considerably lower and higher than the steady-state values at the corresponding pressure during the depressurization respectively. It was suggested that the transitions to stable film boiling at transient critical heat fluxes from intermediate q in in natural convection and from high q in in nucleate boiling for short τ p occur due to explosive-like heterogeneous spontaneous nucleation (HSN). The photographs of typical vapor behavior due to the HSN during depressurization from natural convection regime for short τ p were shown. (author)

Experimental study and modelization of a propane storage tank depressurization

International Nuclear Information System (INIS)

Veneau, Tania

1995-01-01

The risks associated with the fast depressurization of propane storage tanks reveals the importance of the 'source term' determination. This term is directly linked, among others, to the characteristics of the jet developed downstream of the breach. The first aim of this work was to provide an original data bank concerning drop velocity and diameter distributions in a propane jet. For this purpose, a phase Doppler anemometer bas been implemented on an experimental set-up. Propane blowdowns have been performed with different breach sizes and several initial pressures in the storage tank. Drop diameter and velocity distributions have been investigated at different locations in the jet zone. These measurements exhibited the fragmentation and vaporisation trends in the jet. The second aim of this work concerned the 'source term'. lt required to study the coupling between the fluid behaviour inside the tank and the flow through the breach. This model took into account the phase exchange when flashing occurred in the tank. The flow at the breach was described with an homogeneous relaxation model. This coupled modelization has been successfully and exhaustively validated. lt originality lies on the application to propane flows. (author) [fr

An intermediate heat exchanging-depressurizing loop for nuclear hydrogen production

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Soo [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); No, Hee Cheon, E-mail: hcno@kaist.ac.k [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Yoon, Ho Joon; Lee, Jeong Ik [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2010-10-15

Sulfur-iodine (SI) cycle should overcome many engineering challenges to commercialize and prove its feasibilities to compete other thermo-chemical cycles. Some critical issues such as structural material, harsh operating condition and high capital costs were considered obstacles to be actualized. Operating SI cycle at low-pressure is one of the solutions to actualize the cycle. The flash operation with over-azeotropic HI at low pressure does not require temperature and pressure as high as those in the existing methods as well as heating for separation. The operation in low pressure reduces corrosion problems and enables us to use flexible selection of structural material. We devised an intermediate heat exchanging-depressurizing loop to eliminate high operating pressure in the hydrogen side as well as a large pressure difference between the reactor side and the hydrogen side. Molten salts are adequate candidates as working fluids under the high-temperature condition with homogeneous phase during pressure changing process. Using molten salts, 2.20-4.65 MW of pumping work is required to change the pressure from 1 bar to 7 MPa. We selected BeF{sub 2}-containing salts as the possible candidates based on preliminary economic and thermal hydraulic consideration.

Depressurization experiments on a plugged fibrous insulation in a horizontal pressure tube

International Nuclear Information System (INIS)

Lang, H.; Weise, H.J.; Ennen, P.

1977-08-01

Hot gas ducts for high-temperature reactors with a helium turbine are subject to additional operational loads not caused by the gas temperature. They include vibrations, caused by high gas velocities or by the sound fields emitted from the turbine, and stresses, originating from fast, short-time pressure changes. Such pressure changes occur as a rule if the generator coupled with the turbine has to be disconnected from the grid. In order to avoid no-load operation of the turbine a bypass between HP and LP side of the turbine is opened. As a consequence of this measure a sudden pressure drop occurs in the free flow cross-section causing differential pressures within the insulation. As the size of these differential pressures depends on the insulating material, the density of plugging, the kind of internals, and on the position and size of the depressurization borings, the pressure distributions in the insulation were measured on a test tube for the HP channel. (orig./RW) [de

Simulation of Fission Product Liftoff Behavior During Depressurization Transients

International Nuclear Information System (INIS)

Tak, Nam-il; Yoon, Churl; Lee, Sung Nam

2016-01-01

As one of crucial technologies for the NHDD project, the development of the GAMMA-FP code is on-going. The GAMMA-FP code is targeted for fission product transport analysis under accident conditions. A well-known experiment named COMEDIE considered two important phenomena, i.e., fission product plateout and liftoff, for fission product transport within the primary circuit of a prismatic high temperature gas cooled reactor. The accumulated fission products on the structural material via the plateout can be liftoff during a blowdown phase after a pipe break accident. Since the fission product liftoff can increase a radioactivity risk, it is important to predict the amount of fission product liftoff during depressurization accidents. In this work, a model for fission product liftoff is implemented into the GAMMA-FP code and the GAMMA-FP code with the implemented model is validated using the COMEDIE blowdown test data. The results of GAMMA-FP show that the GAMMA-FP code can reliably simulate a pressure transient during blowdown phase after a pipe break accident. In addition, a reasonable amount of fission product liftoff was predicted by the GAMMA-FP code. The maximum difference between the measured and predicted liftoff fraction was less than a factor of 10. More in-depth study is required to increase the accuracy of prediction for a fission product liftoff

Liquid entrainment and off-take at the top of the pressurizer in the case of the actuation of safety depressurization system of APR1400

International Nuclear Information System (INIS)

Kim, Chang Hyun; No, Hee Cheon

2003-01-01

In order to determine the bleed capacity of Safety Depressurization System (SDS) of Advanced Power Reactor 1400 (APR1400) in case of Total Loss of Feed Water (TLOFW), we performed an experimental study of liquid entrainment and liquid off-take from the swelled two-phase mixture surface in a vessel. A total of 208 experimental data on the entrainment and off-take are obtained using a test vessel with the height of 2.0m and the inner diameter of 0.3m having a top break with diameter of 0.05m. Two-phase mixture levels are measured by the ultrasonic sensor within ? .77% with respect to the visual level data. Droplet entrainments are measured and compared with the existing pool entrainment data. The empirical correlation for the onset of off-take is developed in terms of the Froude number (Fr g ) at the break and non-dimensional inception height (h b /d). This correlation shows agreement with the present experimental data within ? 5%. The present off-take quality data is in agreement with Schrock's off-take quality correlation with the r.m.s. error of 15.8%. In the present experiment, droplet entrainment E fg strongly depends upon jg * /h * and is proportional to the 7 th power of jg * /h * in the same way as the off-take data

The Temperature of the Dimethylhydrazine Drops Moving in the Atmosphere after Depressurization of the Fuel Tank Rockets

Directory of Open Access Journals (Sweden)

Bulba Elena

2016-01-01

Full Text Available This work includes the results of the numerical modeling of temperature changes process of the dimethylhydrazine (DMH drops, taking into account the radial temperature gradient in the air after the depressurization of the fuel compartments rockets at high altitude. There is formulated a mathematical model describing the process of DMH drops thermal state modifying when it's moving to the Earth's surface. There is the evaluation of the influence of the characteristic size of heptyl drops on the temperature distribution. It's established that the temperatures of the small size droplets practically completely coincide with the distribution of temperature in the atmosphere at altitudes of up to 40 kilometers.

Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

International Nuclear Information System (INIS)

Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

1982-01-01

The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

Ventilation systems analysis during tornado conditions. Progress report, January--June 1975

International Nuclear Information System (INIS)

Bennett, G.A.; Gregory, W.S.; Smith, P.R.

1975-11-01

The principal concern of this investigation is to develop the capability to simulate the dynamic effects of a tornado depressurization on a ventilation system. The basic formulation and solution of the two-zone series model ventilation subsystem is based on lumped parameter component response equations, the isothermal compression of air, and the conservation of mass. Solutions based on these assumptions are also presented for the two-zone series model with natural bypass, the two-zone series model with recirculation, and the natural branching model. A parameter study is presented comparing the effects of changes in system resistance, system capacitance, and variable tornado depressurization rates. The adaptability of the basic formulation to adiabatic compression of air and the addition of duct resistance is examined. A quasi-steady formulation is introduced and preliminary considerations of the importance of inertia are presented. Preliminary conclusions in this area indicate that inertial effects can be neglected. For relatively long ducts slow shock development appears possible. Work on the effect of tornado depressurization rates as related to shock development and on the importance of inertia effects is continuing

Culturable prokaryotic diversity of deep, gas hydrate sediments: first use of a continuous high-pressure, anaerobic, enrichment and isolation system for subseafloor sediments (DeepIsoBUG)

OpenAIRE

Parkes, R John; Sellek, Gerard; Webster, Gordon; Martin, Derek; Anders, Erik; Weightman, Andrew J; Sass, Henrik

2009-01-01

Deep subseafloor sediments may contain depressurization-sensitive, anaerobic, piezophilic prokaryotes. To test this we developed the DeepIsoBUG system, which when coupled with the HYACINTH pressure-retaining drilling and core storage system and the PRESS core cutting and processing system, enables deep sediments to be handled without depressurization (up to 25 MPa) and anaerobic prokaryotic enrichments and isolation to be conducted up to 100 MPa. Here, we describe the system and its first use...

Hydrate prevention in petroleum production sub sea system

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Paula L.F.; Rocha, Humberto A.R. [Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Rodrigues, Antonio P. [Universidade Federal do Rio Grande do Norte (UFRN), Natal, RN (Brazil)

2012-07-01

In spite of the merits of the several hydrate prevention techniques used nowadays, such as: chemical product injection for inhibition and use of thick thermal insulate lines; hydrates per times happen and they are responsible for considerable production losses. Depressurization techniques can be used so much for prevention as in the remediation. Some hydrate removal techniques need a rig or vessel, resources not readily available and with high cost, reason that limits such techniques just for remediation and not for prevention. In the present work it is proposed and described an innovative depressurization system, remote and resident, for hydrate prevention and removal, applicable as for individual sub sea wells as for grouped wells by manifold. Based on low cost jet pumps, without movable parts and with a high reliability, this technique allows hydrate prevention or remediation in a fast and remote way, operated from the production unit. The power fluid line and fluid return line can be integrated in the same umbilical or annulus line structure, without significant increase in the construction costs and installation. It is not necessary to wait for expensive resource mobilization, sometimes not available quickly, such as: vessels or rigs. It still reduces the chemical product consumption and permits to depressurized stopped lines. Other additional advantage, depressurization procedure can be used in the well starting, removing fluid until riser emptying. (author)

Mudstone depressurization behaviour in an open pit coal mine, Indonesia

Energy Technology Data Exchange (ETDEWEB)

Marchand, G.; Waterhouse, J. [Golder Associates, West Perth, WA (Australia); Crisostomo, J. [PT Adaro Indonesia, Jakarta (Indonesia)

2010-07-01

Mining activities in the Tutupan mine in Indonesia began in the mid-1990s. The open pit mine's coal seams are interbedded with fine-grained sandstones, mudstones, and carbonaceous mudstones. Slope stability analyses at the pit have integrated hydrogeology with geotechnical engineering analyses to optimize slope designs and reduce the risk of slope failure. This paper discussed the impact of mining and dewatering on mudstone depressurization. Sensors were placed at key points in the mine to obtain data related to the mudstone units. Reductions in pore pressure occurred as a result of groundwater flow away from the observed zones, increases in porosity, and increases in total porosity caused by an expansion of the rock mass as a result of drainage and hydrostatic unloading. Mudstone pore pressure trends with time were interpreted by determining the thickness of the mudstone unit, the presence or absence of known thin sandstone beds, unloading from overhead mining activities, and the position of the mudstone within the sedimentary sequence. The study showed that unloading activities have a significant impact on pore pressure in thick mudstone units, regardless of the depth, thickness, or properties of the unit. Pore pressure within high wall mudstone units typically decreased to values equivalent to the elevation of the unit where it was exposed to dips in a high wall. The dewatering of sandstone units in low walls caused a decline in pore pressure within the thick mudstone units located beneath the sandstones. Differences in primary permeabilities were attributed to greater fracturing in deeper and stronger rock units. 3 refs., 4 figs.

Numerical simulation of Class 3 hydrate reservoirs exploiting using horizontal well by depressurization and thermal co-stimulation

International Nuclear Information System (INIS)

Yang, Shengwen; Lang, Xuemei; Wang, Yanhong; Wen, Yonggang; Fan, Shuanshi

2014-01-01

Highlights: • Depressurization and thermal co-stimulation using horizontal well were proposed. • 3D stimulation showed that gas release rate was 3 × 10 5 m 3 per day within 450 days. • 2D stimulation showed that Class 3 hydrates could be dissociated within 8500 days. • 2D Simulation showed that heat flow was 1620 W lasting 1500 days, and decreased fast. • 1.1× 10 5 kg water was collected within 2000 days and then no more water was produced. - Abstract: Class 3 hydrate reservoirs exploiting using horizontal well by depressurization and thermal co-stimulation was simulated using the HydarteResSim code. Results showed that more than 20% of hydrates in the reservoirs had been dissociated within 450 days at the well temperature of 42 °C and well pressure of 0.1P 0 , 0.2P 0 (P 0 is the initial pressure of the reservoirs, simplifying 42 °C and 0.1P 0 , 42 °C and 0.2P 0 ). While the production behavior of 42 °C and 0.5P 0 , 42 °C and 0.8P 0 were not so exciting. In order to understand the production character of the well in long term, the cross section of 1 m length reservoirs was simulated. Simulation results showed that 4.5 × 10 5 m 3 gas would be collected within 4500 days and 1.1 × 10 6 kg water could be produced within 1500 days in the well at 42 °C and 0.1P 0 . 3.5 × 10 5 m 3 gas would be collected within 8500 days and 1.1 × 10 6 kg water could be produced within 1500 days in the well at 42 °C and 0.2P 0 . The heat flow was 1620 W at the beginning and then decreased rapidly in the two cases. For reservoirs of 1495.2 m in length, about 6.7 × 10 8 m 3 and 5.3 × 10 8 m 3 gas would be collected in the well corresponding to conditions of 42 °C and 0.1P 0 , and 42 °C and 0.2P 0

A concept of JAERI passive safety light water reactor system (JPSR)

Energy Technology Data Exchange (ETDEWEB)

Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

1995-09-01

The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

Assessment of Literature Related to Combustion Appliance Venting Systems

Energy Technology Data Exchange (ETDEWEB)

Rapp, Vi H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Singer, Brett C. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Stratton, Chris [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Wray, Craig P. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2012-06-01

In many residential building retrofit programs, air tightening to increase energy efficiency is constrained by concerns about related impacts on the safety of naturally vented combustion appliances. Tighter housing units more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spillage. Several test methods purportedly assess the potential for depressurization-induced backdrafting and spillage, but these tests are not robustly reliable and repeatable predictors of venting performance, in part because they do not fully capture weather effects on venting performance. The purpose of this literature review is to investigate combustion safety diagnostics in existing codes, standards, and guidelines related to combustion appliances. This review summarizes existing combustion safety test methods, evaluations of these test methods, and also discusses research related to wind effects and the simulation of vent system performance. Current codes and standards related to combustion appliance installation provide little information on assessing backdrafting or spillage potential. A substantial amount of research has been conducted to assess combustion appliance backdrafting and spillage test methods, but primarily focuses on comparing short-term (stress) induced tests and monitoring results. Monitoring, typically performed over one week, indicated that combinations of environmental and house operation characteristics most conducive to combustion spillage were rare. Research, to an extent, has assessed existing combustion safety diagnostics for house depressurization, but the objectives of the diagnostics, both stress and monitoring, are not clearly defined. More research is also needed to quantify the frequency of test “failure” occurrence throughout the building stock and assess the statistical effects of weather (especially wind) on house depressurization and in turn on combustion appliance venting

Recent Ground Hold and Rapid Depressurization Testing of Multilayer Systems

Science.gov (United States)

Johnson, Wesley L.

2014-01-01

In the development of flight insulation systems for large cryogenic orbital storage (spray on foam and multilayer insulation), testing need include all environments that are experienced during flight. While large efforts have been expended on studying, bounding, and modeling the orbital performance of the insulation systems, little effort has been expended on the ground hold and ascent phases of a mission. Historical cryogenic in-space systems that have flown have been able to ignore these phases of flight due to the insulation system being within a vacuum jacket. In the development phase of the Nuclear Mars Vehicle and the Shuttle Nuclear Vehicle, several insulation systems were evaluated for the full mission cycle. Since that time there had been minimal work on these phases of flight until the Constellation program began investigating cryogenic service modules and long duration upper stages. With the inception of the Cryogenic Propellant Storage and Transfer Technology Demonstration Mission, a specific need was seen for the data and as such, several tests were added to the Cryogenic Boil-off Reduction System liquid hydrogen test matrix to provide more data on a insulation system. Testing was attempted with both gaseous nitrogen (GN2) and gaseous helium (GHe) backfills. The initial tests with nitrogen backfill were not successfully completed due to nitrogen liquefaction and solidification preventing the rapid pumpdown of the vacuum chamber. Subsequent helium backfill tests were successful and showed minimal degradation. The results are compared to the historical data.

System and method for continuous solids slurry depressurization

Science.gov (United States)

Leininger, Thomas Frederick; Steele, Raymond Douglas; Cordes, Stephen Michael

2017-07-11

A system includes a first pump having a first outlet and a first inlet, and a controller. The first pump is configured to continuously receive a flow of a slurry into the first outlet at a first pressure and to continuously discharge the flow of the slurry from the first inlet at a second pressure less than the first pressure. The controller is configured to control a first speed of the first pump against the flow of the slurry based at least in part on the first pressure, wherein the first speed of the first pump is configured to resist a backflow of the slurry from the first outlet to the first inlet.

System transient response to loss of off-site power

International Nuclear Information System (INIS)

Sozer, A.

1990-01-01

A simultaneous trip of the reactor, main circulation pumps, secondary coolant pumps, and pressurizer pump due to loss of off-site power at the High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory (ORNL) has been analyzed to estimate available safety margin. A computer model based on the Modular Modeling System code has been used to calculate the transient response of the system. The reactor depressurizes from 482.7 psia down to about 23 psia in about 50 seconds and remains stable thereafter. Available safety margin has been estimated in terms of the incipient boiling heat flux ratio. It is a conservative estimate due to assumed less than available primary and secondary flows and higher than normal depressurization rate. The ratio indicates no incipient boiling conditions at the hot spot. No potential damage to the fuel is likely to occur during this transient. 2 refs., 6 figs

System and method for continuous solids slurry depressurization

Science.gov (United States)

Leininger, Thomas Frederick; Steele, Raymond Douglas; Yen, Hsien-Chin William; Cordes, Stephen Michael

2017-10-10

A continuous slag processing system includes a rotating parallel disc pump, coupled to a motor and a brake. The rotating parallel disc pump includes opposing discs coupled to a shaft, an outlet configured to continuously receive a fluid at a first pressure, and an inlet configured to continuously discharge the fluid at a second pressure less than the first pressure. The rotating parallel disc pump is configurable in a reverse-acting pump mode and a letdown turbine mode. The motor is configured to drive the opposing discs about the shaft and against a flow of the fluid to control a difference between the first pressure and the second pressure in the reverse-acting pump mode. The brake is configured to resist rotation of the opposing discs about the shaft to control the difference between the first pressure and the second pressure in the letdown turbine mode.

Depressurization accident analysis of MPBR by PBRSIM with chemical reaction model

International Nuclear Information System (INIS)

No, Hee Cheon; Kadak, A. C.

2002-01-01

The simple model for natural circulation is implemented into PBR S IM to provide air inlet velocity from the containment air space. For the friction and form loss only the pebble region is considered conservatively modeling laminar flow through a packed bed. For the chemical reaction model of PBR S IM the oxidation rate is determined as the minimum value of three mechanisms estimated at each time step: oxygen mass flow rate entering the bottom of the reflector, oxidation rate by kinetics, and oxygen mass flow rate arriving at the graphite surface by diffusion. Oxygen mass flux arriving at the graphite surface by diffusion is estimated based on energy-mass analogy. Two types of exothermic chemical reaction are considered: (C + zO 2 → xCO + yCO 2 ) and (2CO + O 2 2CO 2 ). The heterogeneous and homogeneous chemical reaction rates by kinetics are determined by INEEL and Bruno correlations, respectively. The instantaneous depressurization accident of MPBR is simulated using PBR S IM with chemical model. The air inlet velocity is initially rapidly dropped within 10 hr and reaches a saturation value of about 1.5cm/s. The oxidation rate by the diffusion process becomes lower than that by the chemical kinetics above 600K. The maximum pebble bed temperatures without and with chemical reaction reach the peak values of 1560 and 1617 .deg. C at 80 hr and 92 hr, respectively. As the averaged temperatures in the bottom reflector and the pebble bed regions increase with time, (C+1/2O2 ->CO) reaction becomes dominant over (C+O 2 →CO 2 ) reaction. Also, the CO generated by (C+1/2O 2 →CO) reaction will be consumed by (2CO+O 2 →2CO 2 ) reaction and the energy homogeneously generated by this CO depletion reaction becomes dominant over the heterogeneous reaction

CNTB program for the analysis of partially mixed containment atmospheres during depressurization events

International Nuclear Information System (INIS)

Landoni, J.A.

1979-07-01

This program describes the analytical models for the CNTB computer program, which is permanently filed in the archive library of the General Atomic (GA) San Diego Data Center under reference number THSD-2699. Developed during the last four years, this computer program has been successfully applied in its presented form to the type of containment atmosphere transients illustrated in this report. For example, the CNTB computer program is applicable (1) to the design basis depressurization accident (DBDA) to determine the effect of the partial mixing on the containment atmospheric peak pressure (known as nonmixing penalty) and (2) for Class 9 accidents, such as the loss of forced circulation (LOFC), for the AIPA Phase I studies. The capability of the CNTB computer program has been substantially improved over its precursor, the CONTEMPT-G computer program, to predict the thermodynamic behavior of the containment atmosphere during helium releases, assuming partial mixing of the original air with the effluent and to predict the amount of the environmental leaks under closed and open containment conditions. In addition, the CNTB computer program running times are considerably below the ones required for the CONTEMPT-G computer program. Computational solution of the variable parameters in the containment atmosphere is effected by an iterative technique, while the temperatures for its boundaries are obtained by finite differences. The CNTB computer program, written in FORTRAN V, has been implemented at GA on the UNIVAC 1110 computer

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

POSSIBLE ROLE OF INDOOR RADON REDUCTION SYSTEMS IN BACK-DRAFTING RESIDENTIAL COMBUSTION APPLIANCES

Science.gov (United States)

The article gives results of a computational sensitivity analysis conducted to identify conditions under which residential active soil depressurization (ASD) systems for indoor radon reduction might contribute to or create back-drafting of natural draft combustion appliances. Par...

Radon mitigation experience in houses with basements and adjoining crawl spaces

International Nuclear Information System (INIS)

Messing, M.; Henschel, D.B.

1990-01-01

Active soil depressurization systems were installed in four basement houses with adjoining crawl spaces in Maryland. In addition, existing soil depressurization systems were modified in two additional basement-plus-crawl-space houses. These six houses were selected to include both good and poor communication beneath the basement slab, and different degrees of importance of the crawl space as a source of the indoor radon. The radon reduction effectiveness was compared for: depressurization only under the basement slab; depressurization only under a polyethylene liner over the unpaved crawl-space floor; and simultaneous depressurization under both the basement slab and the crawl-space liner. The objective of this paper is to identify under what conditions treatment of the basement alone might provide sufficient radon reductions in houses of this substructure, and what incremental benefits might be achieved by also treating the crawl space

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.

Scaling relation and regime map of explosive gas–liquid flow of binary Lennard-Jones particle system

KAUST Repository

Inaoka, Hajime; Yukawa, Satoshi; Ito, Nobuyasu

2012-01-01

liquid droplets, and gas particles, which remain supercritical gaseous states under the depressurization realized by simulations. The system has a pipe-like structure similar to the model of a shock tube. We observed physical quantities and flow regimes

Thermo-hydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

International Nuclear Information System (INIS)

Abouhadra, D.S.; Byrne, J.E.

2003-01-01

In nuclear reactors of the magnox or advanced gas cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accidents using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear electric . The tests were carried out on the thermal hydraulics experimental research assembly (THERA) loop at manchester university. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 k, 50 k, and 100 k was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s

Assessment of Literature Related to Combustion Appliance Venting Systems

Energy Technology Data Exchange (ETDEWEB)

Rapp, V. H.; Less, B. D.; Singer, B. C.; Stratton, J. C.; Wray, C. P.

2015-02-01

In many residential building retrofit programs, air tightening to increase energy efficiency is often constrained by safety concerns with naturally vented combustion appliances. Tighter residential buildings more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spill combustion exhaust into the living space. Several measures, such as installation guidelines, vent sizing codes, and combustion safety diagnostics, are in place with the intent to prevent backdrafting and combustion spillage, but the diagnostics conflict and the risk mitigation objective is inconsistent. This literature review summarizes the metrics and diagnostics used to assess combustion safety, documents their technical basis, and investigates their risk mitigations. It compiles information from the following: codes for combustion appliance venting and installation; standards and guidelines for combustion safety diagnostics; research evaluating combustion safety diagnostics; research investigating wind effects on building depressurization and venting; and software for simulating vent system performance.

Data report of ROSA/LSTF experiment SB-HL-12. 1% hot leg break LOCA with SG depressurization and gas inflow

International Nuclear Information System (INIS)

Takeda, Takeshi

2016-01-01

An experiment SB-HL-12 was conducted on February 24, 1998 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-HL-12 simulated a 1% hot leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). Steam generator (SG) secondary-side depressurization by fully opening the relief valves in both SGs as an accident management (AM) action was initiated immediately after maximum surface temperature of simulated fuel rod reached 600 K. Auxiliary feedwater injection into the secondary-side of both SGs was started immediately after the initiation of AM action. After the onset of AM action due to first core uncovery by core boil-off, the primary pressure decreased following the SG secondary-side pressure, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before loop seal clearing (LSC) induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after the LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after the ACC tanks started to discharge nitrogen gas, which resulted in no actuation of LPI system of ECCS during the experiment. Third core uncovery by core boil-off occurred during the reflux condensation in the SG U-tubes under nitrogen gas inflow. The core power was automatically decreased by the LSTF core protection system when the maximum fuel rod surface temperature exceeded 908 K. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal

Development of technique for estimating primary cooling system break diameter in predicting nuclear emergency event sequence

International Nuclear Information System (INIS)

Tatebe, Yasumasa; Yoshida, Yoshitaka

2012-01-01

If an emergency event occurs in a nuclear power plant, appropriate action is selected and taken in accordance with the plant status, which changes from time to time, in order to prevent escalation and mitigate the event consequences. It is thus important to predict the event sequence and identify the plant behavior resulting from the action taken. In predicting the event sequence during a loss-of-coolant accident (LOCA), it is necessary to estimate break diameter. The conventional method for this estimation is time-consuming, since it involves multiple sensitivity analyses to determine the break diameter that is consistent with the plant behavior. To speed up the process of predicting the nuclear emergency event sequence, a new break diameter estimation technique that is applicable to pressurized water reactors was developed in this study. This technique enables the estimation of break diameter using the plant data sent from the safety parameter display system (SPDS), with focus on the depressurization rate in the reactor cooling system (RCS) during LOCA. The results of LOCA analysis, performed by varying the break diameter using the MAAP4 and RELAP5/MOD3.2 codes, confirmed that the RCS depressurization rate could be expressed by the log linear function of break diameter, except in the case of a small leak, in which RCS depressurization is affected by the coolant charging system and the high-pressure injection system. A correlation equation for break diameter estimation was developed from this function and tested for accuracy. Testing verified that the correlation equation could estimate break diameter accurately within an error of approximately 16%, even if the leak increases gradually, changing the plant status. (author)

Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

International Nuclear Information System (INIS)

Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

2012-06-01

The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

ECCS control circuit

International Nuclear Information System (INIS)

Sato, Takashi.

1986-01-01

Purpose: To afford a sufficient margin to pressure vibrations upon starting of an automatic depressurization system by dispersing pressure vibration in suppression water due to the opening action of an automatic releaf valve in the automatic depressurization system thereby reducing the dynamic load exerted to the surface of the suppression walls. Constitution: Upon occurrence of loss of coolant accidents, an automatic releaf valve for automatic depressurization is opened to deliver the steams in the pressure vessel into the suppression pool. Since a plurality of automatic releaf valves have usually been disposed, if they are opened simultaneously, excess dynamic loads are exerted due to the pressure vibrations to the wall surface of the suppression pool. In this invention, a control circuit is disposed such that the opening timing for each of the automatic releaf valves is deviated upon occurrence of a driving signal for the automatic depressurization system to thereby disperse the pressure vibrations in the suppression water. (Kamimura, M.)

Pipe stress intensity factors and coupled depressurization and dynamic crack propagation. 1976 Annual report

International Nuclear Information System (INIS)

Emery, A.F.; Kobayashi, A.S.; Love, W.J.

1978-04-01

This report contains the description of predictive models for the initiation and propagation of cracks in pipes and the numerical results obtained. The initiation of the crack was studied by evaluating stress intensity factors under static conditions for a series of representative flaws. Three-dimensional static stress intensity factors were determined for quarter-elliptical cracks at the corner of a hole in an infinite plate and at the corner of a bore in a rotating disk. Semi-elliptical cracks for plates in bending and in pressurized and thermally stressed hollow cylinders were also evaluated. The stress fields, in the absence of a crack, were used in the ''alternating technique'' to compute the stress intensity factors along the crack front. Parametric studies were made to assess the effects of crack thickness, the ratio of the major and minor axes of the ellipse and the thickness of the cylinders or plates. These parametric results may be used to predict critical flaw sizes for the initiation of the running crack. The initiation and propagation of axial through cracks in pressurized pipes was studied by using an elastic-plastic finite different shell code coupled with a one-dimensional thermal-hydraulic code which computed the leakage through the crack opening and the depressurization of the fluid in the pipe. The effects of large deflections and different fluid pressure profiles were investigated. The results showed that the crack opening shape is dependent upon the fracture criterion used and upon the average pressure on the crack flaps, but not upon the specific pressure profile. The consideration of large deflections changed the opening size of the crack and through the coupling with the pipe pressures, strongly affected the crack tip speed. However, for equal crack lengths, there was little difference between calculations made for large and small deflection

First use of ECOSIM in air management system

International Nuclear Information System (INIS)

Perez Vara, R.; Torroglosa Ponce, V.; Lebru, A.; Novara, M.

1993-01-01

ECOSIM is a software tool for the simulation of Environmental Control and Life Support (ECLS) systems which has been developed for the European Space Agency. A preliminary model of the Hermes Air Management System has been developed during the ECOSIM testing in order to assess the functionality of the software and to verify its results with those obtained from previous simulation tools. The model represents the Hermes cabin with its crew and it includes submodels for the sub-systems performing the following functions: - Temperature and Humidity Control; - Total Pressure and Composition Control; - Air revitalisation. The interactions between these different subsystem are taken into account by the model, while many of the previous simulations made assumptions to decouple the different subsystems (e.g: a constant cabin temperature has been assumed during cabin depressurization transients, to decouple the pressure control section from the air conditioning section). The model allows to check the validity of these kind of assumption. The paper also presents the results obtained with this model for the following experiments: - Normal 24 hour cycle; - Re-entry; - Depressurization due to a micro-meteorite hole; - Failure of the condensate water separators. The first version of ECOSIM will be released to the space industry before the end of year 1992 and the user's feedback will lead to improvements being implemented in later versions. (author)

Assessment of Gas Production Potential from Hydrate Reservoir in Qilian Mountain Permafrost Using Five-Spot Horizontal Well System

Directory of Open Access Journals (Sweden)

Yun-Pei Liang

2015-09-01

Full Text Available The main purpose of this study is to investigate the production behaviors of gas hydrate at site DK-2 in the Qilian Mountain permafrost using the novel five-spot well (5S system by means of numerical simulation. The whole system is composed of several identical units, and each single unit consists of one injection well and four production wells. All the wells are placed horizontally in the hydrate deposit. The combination method of depressurization and thermal stimulation is employed for hydrate dissociation in the system. Simulation results show that favorable gas production and hydrate dissociation rates, gas-to-water ratio, and energy ratio can be acquired using this kind of multi-well system under suitable heat injection and depressurization driving forces, and the water production rate is manageable in the entire production process under current technology. In addition, another two kinds of two-spot well (2S systems have also been employed for comparison. It is found that the 5S system will be more commercially profitable than the 2S configurations for gas production under the same operation conditions. Sensitivity analysis indicates that the gas production performance is dependent on the heat injection rate and the well spacing of the 5S system.

Follow-up durability measurements and mitigation-performance improvement tests in 38 Eastern Pennsylvania houses having indoor radon-reduction systems. Final report, Oct 89-Feb 90

International Nuclear Information System (INIS)

Findlay, W.O.; Robertson, A.; Scott, A.G.

1991-03-01

The report gives results of follow-up tests in 38 difficult-to-mitigate Pennsylvania houses where indoor radon reduction systems had been installed 2 to 4 years earlier. Objectives were to assess system durability, methods for improving performance, and methods for reducing installation and operating costs. The durability tests indicated that the 38 systems have not experienced any significant degradation in indoor radon levels or in system flows/suctions, except in 6 houses where system fans failed, and in houses where homeowners turned off the systems. Tests to improve performance indicated that nearly all of the elevated residual radon levels are due to re-entrainment back into the house of very-high-radon exhaust gas from the soil depressurization systems, and to radon release from well water. Tests to reduce system costs showed that premitigation sub-slab suction field measurements can help prevent installation of too many suction pipes when communication is good, but suggest a need for too many pipes when communication is poor. Soil depressurization fans could not be turned down to the extent expected in some systems that were over-designed. Between 6 and 42% of the exhausted air was withdrawn from the house

Concept Study for Military Port Design Using Natural Processes.

Science.gov (United States)

1982-06-15

concept, which he called pressure retarded osmosis. In his system "the volume-enhanced brine would be subsequently depressurized through a hydroturbine ...pressure gradient, i.e. the flux is "uphill". The subsequent depressurization of the permeate through a hydroturbine -generator set would produce

Analysis of an ADS spurious opening event at a BWR/6 by means of the TRACE code

International Nuclear Information System (INIS)

Nikitin, Konstantin; Manera, Annalisa

2011-01-01

Highlights: → The spurious opening of 8 relief valves of the ADS system in a BWR/6 has been simulated. → The valves opening results in a fast depressurization and significant loads on the RPV internals. → This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the available plant data. - Abstract: The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.

Apparatus and method for depressurizing, degassing, and affording decay of the radioactivity of weakly radioactive condensates in nuclear power plants

International Nuclear Information System (INIS)

Gross, R.; Plotz, J.

1976-01-01

Described is an apparatus for depressurizing, degassing and affording decay of weakly radioactive condensates in nuclear power plants having a turbine and a main condenser turbine wherein exhaust steam of the turbine is condensed and forms a main condensate, and includes a collecting tank for the condensate situated below the condenser. A plurality of horizontal degassing channels, each having a lateral overflow, are disposed in the upper part of the condensate collecting tank and are filled with the main condensate up to the level of the overflow. At least one feedwater preheater which is heated by bleeder steam from the turbine provides a secondary condensate. Below the overflow height of the degassing channels extend horizontal feed pipes for the secondary condensate. The feed pipes are connected to the output of pressure relieving expanding devices and are provided on their underside with discharge openings for the bubbling of the secondary condensate into the main condensate to thereby degass the main condensate. The condensate collecting tank has mutually offset partitions therein providing an adequately long path for the decay of the main and secondary condensates. The condensate which is discharged from the condensate collecting tank is returned into the cycle as feedwater. Also disclosed is a method of operating the foregoing apparatus

Drive transmission system between a driving organ and a receiver organ

International Nuclear Information System (INIS)

Guillot, J.F.

1985-01-01

The present invention applies to the control rods of a water cooled nuclear reactor. The drive transmission system is disposed on the internal kinematic chain, between the control rod which is the receiver organ, and the driving organ. The control rod translation is obtained from a motion of rotation transformed in a motion of translation by means of a screw-nut system. The present invention prevents from control rod ejection in case of depressurization of the vessel containing the control rod drives or in case of reactor upsetting when it is embarked [fr

Rehabilitation of a house with high radon level, using a ground ventilation system with double barrier

International Nuclear Information System (INIS)

Bonnefous, Y.C.; Richon, P.; Arnautou, J.C.; Sabroux, J.C.

1995-01-01

A ground ventilation system has been designed and implemented in a town hall in Brittany. Radon concentration in the heating unit room of this building has been reduced from 10000 Bq/m 3 to less than 200 Bq/m 3 by the means of a depressurization system using a 32 W fan, which blows air into a permeable gravel layer intercalated between two radon barrier mylar films. Results show that passive systems should be applicable; for new buildings, very low energy consumption systems with 10 W fans, are easily implemented if designed before construction

Alternative protections for loss of coolant accidents

International Nuclear Information System (INIS)

Estevez, E.A.

1997-01-01

One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

Impact of ventilation systems and energy savings in a building on the mechanisms governing the indoor radon activity concentration.

Science.gov (United States)

Collignan, Bernard; Powaga, Emilie

2017-11-23

For a given radon potential in the ground and a given building, the parameters affecting the indoor radon activity concentration (IRnAC) are indoor depressurization of a building and its air change rate. These parameters depend mainly on the building characteristics, such as airtightness, and on the nature and performances of the ventilation system. This study involves a numerical sensitivity assessment of the indoor environmental conditions on the IRnAC in buildings. A numerical ventilation model has been adapted to take into account the effects of variations in the indoor environmental conditions (depressurization and air change rate) on the radon entry rate and on the IRnAC. In the context of the development of a policy to reduce energy consumption in a building, the results obtained showed that IRnAC could be strongly affected by variations in the air permeability of the building associated with the ventilation regime. Copyright © 2017 Elsevier Ltd. All rights reserved.

Current Status of Aerosol Generation and Measurement Facilities for the Verification Test of Containment Filtered Venting System in KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Il; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, the design of aerosol generation and measurement systems are explained and present circumstances are also described. In addition, the aerosol test plan is shown. Containment Filtered Venting System (FCVS) is one of the safety features to reduce the amount of released fission product into the environment by depressurizing the containment. Since Chernobyl accident, the regulatory agency in several countries in Europe such as France, Germany, Sweden, etc. have been demanded the installation of the CFVS. Moreover, the feasibility study on the CFVS was also performed in U.S. After the Fukushima accident, there is a need to improve a containment venting or installation of depressurizing facility in Korea. As a part of a Ministry of Trade, Industry and Energy (MOTIE) project, KAERI has been conducted the integrated performance verification test of CFVS. As a part of the test, aerosol generation system and measurement systems were designed to simulate the fission products behavior. To perform the integrated verification test of CFVS, aerosol generation and measurement system was designed and manufactured. The component operating condition is determined to consider the severe accident condition. The test will be performed in normal conditions at first, and will be conducted under severe condition, high pressure and high temperature. Undesirable difficulties which disturb the elaborate test are expected, such as thermophoresis on the pipe, vapor condensation on aerosol, etc.

Scaling relation and regime map of explosive gas–liquid flow of binary Lennard-Jones particle system

KAUST Repository

Inaoka, Hajime

2012-02-01

We study explosive gasliquid flows caused by rapid depressurization using a molecular dynamics model of Lennard-Jones particle systems. A unique feature of our model is that it consists of two types of particles: liquid particles, which tend to form liquid droplets, and gas particles, which remain supercritical gaseous states under the depressurization realized by simulations. The system has a pipe-like structure similar to the model of a shock tube. We observed physical quantities and flow regimes in systems with various combinations of initial particle number densities and initial temperatures. It is observed that a physical quantity Q, such as pressure, at position z measured along a pipe-like system at time t follows a scaling relation Q(z,t)=Q(zt) with a scaling function Q(ζ). A similar scaling relation holds for time evolution of flow regimes in a system. These scaling relations lead to a regime map of explosive flows in parameter spaces of local physical quantities. The validity of the scaling relations of physical quantities means that physics of equilibrium systems, such as an equation of state, is applicable to explosive flows in our simulations, though the explosive flows involve highly nonequilibrium processes. In other words, if the breaking of the scaling relations is observed, it means that the explosive flows cannot be fully described by physics of equilibrium systems. We show the possibility of breaking of the scaling relations and discuss its implications in the last section. © 2011 Elsevier B.V. All rights reserved.

Inactivation of Staphylococcus aureus in raw salmon with supercritical CO2 using experimental design

Directory of Open Access Journals (Sweden)

Mônica CUPPINI

2016-01-01

Full Text Available Abstract Considering the microbial safety of consumption of raw foods (Asian food, this study aimed to explore the inactivation S. aureus in raw salmon by supercritical CO2 treatment (SC-CO2. For this purpose, experimental design methodology was employed as a tool to evaluate the effects of pressure (120-220 bar, the depressurization rate (10 to 100 bar.min–1 and the salmon:CO2 mass relation (1:0.2 to 1:1.0. It was observed that the pressure and the depressurization rate was statistically significant, i.e. the higher the system pressure and depressurization rate, the greater the microbial inactivation. The salmon: CO2 mass relation did not influence the S. aureus inactivation in raw salmon. There was a total reduction in S. aureus with 225 bar, a depressurizing rate of 100 bar.min–1, a salmon: CO2 mass relation of 1:0.6, for 2 hours at 33 °C.

Thermal-hydraulic tests for reactor safety system

International Nuclear Information System (INIS)

Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

2002-05-01

Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

Fission product plateout and liftoff in the MHTGR primary system: A review

International Nuclear Information System (INIS)

Wichner, R.P.

1991-04-01

A review is presented of the technical basis for predicting radioactivity release resulting from depressurization of an MHTGR primary system. Consideration is restricted to so called dry events with no involvement of the steam system. The various types of deposition mechanisms effective for iodine, cesium, strontium, and silver are discussed in terms of their chemical characteristics and the nature of the materials in the primary system. Emphasis is given to iodine behavior, including means for estimating the quantity available for release, the types of plateout locations in the primary system, and the effect of dust on distribution and release. The behavior of fission products cesium, strontium, and silver in such accidents is presented qualitatively. A major part of the review deals with expected dust levels, types, and transport. Available information on the level and nature of dust in the HTGR primary system is reviewed. A summary is presented of dust deposition and liftoff mechanisms. It was concluded that recent approaches to dust liftoff modeling, based on turbulent burst concepts for removal from surfaces, probably offer advantages over the current shear ratio approach. This study concludes that iodine releases from dry depressurization events are likely to be extremely low, on the order of millicuries, due to a predictably low degree of chemical desorption, a low degree of dust liftoff, and a low involvement of iodine with dust. It was also concluded that deposition mechanisms controlling the distribution of fission product material in the primary system, and hence also controlling the degree of liftoff, depend strongly on the chemical nature of the individual elements. Therefore contrary to the current practice, both plateout and liftoff models should reflect those unique chemical and physical properties. 56 refs., 16 figs., 23 tabs

CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application

International Nuclear Information System (INIS)

Mankamo, T.; Bjoere, S.; Olsson, Lena

1992-12-01

Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems

The establishment and analysis of TRACE model for ultimate response guideline of Chinshan nuclear power plant - 15448

International Nuclear Information System (INIS)

Huang, J.J.; Wang, J.R.; Shih, C.; Chen, S.W.; Liao, L.Y.; Lin, H.T.

2015-01-01

The purpose of this research is to use TRACE code to perform a simulation that executes the procedures of URG (Ultimate Response Guidelines) to deal with Fukushima-like accidents. TRACE is an advanced thermal hydraulic code that has been developed by the United States Nuclear Regulatory Commission for NPP safety analysis. In this work TRACE has been used to analyze the thermal hydraulic model for the URG of the Chinshan nuclear power plant that is composed of 2 BWR-type reactors. URG includes 2-stage depressurization, alternative water injection and removing decay heat through the ejection from containment. The 2-stage depressurization strategy includes controlled depressurization and emergency depressurization to replace traditional one-stage depressurization. Results show that by comparing with one-stage depressurization strategy, 2-stage depressurization strategy is able to reduce peak cladding temperature (PCT) effectively and needs much less minimum flow rate of alternative water injection in the accident

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

International Nuclear Information System (INIS)

1976-09-01

Thermo-hydraulic behavior in the primary cooling system of a pressurized water reactor with an upper head injection system (UHI) in a postulated loss-of-coolant accident (LOCA) has been studied with ROSA-II test facility. Simulated UHI and internal structures of the pressure vessel were installed to the facility for the experiment. Nine maximum-sized double-ended break tests and one medium-sized split break test were performed for the cold-leg break condition. The results are as follows: (1) Fluid mixing in the upper head is not perfect. (2) Cold water injection into the steam or two-phase fluid causes violent depressurization due to the condensation. Flow pattern in the primary cooling system is largely influenced by the above two. (auth.)

Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

International Nuclear Information System (INIS)

Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

2009-01-01

The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

Energy Technology Data Exchange (ETDEWEB)

Wu, Wen-Hsiung, E-mail: whwu1127@aec.gov.tw [Atomic Energy Council, 2F., No. 80, Sec.1, Chenggong Rd., Yonghe Dist., New Taipei City 234, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Guangfu Rd., Hsinchu City 300, Taiwan (China); Liao, Lih-Yih, E-mail: lyliao@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 325, Taiwan (China)

2016-12-01

Highlights: • Discuss the problem of EOPs at the time of Fukushima accident to deal with the prolonged SBO. • Elaborate the potential risk accompanied with the emergency depressurization in the SBO. • Describe a special guideline to cope with Fukushima-like accidents and provide its technical basis. • Point out that Fukushima accident might have been prevented if improved EOPs had been used. • Propose key points and suggestions for improving the EOPs. - Abstract: One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima

Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

International Nuclear Information System (INIS)

Wu, Wen-Hsiung; Liao, Lih-Yih

2016-01-01

Highlights: • Discuss the problem of EOPs at the time of Fukushima accident to deal with the prolonged SBO. • Elaborate the potential risk accompanied with the emergency depressurization in the SBO. • Describe a special guideline to cope with Fukushima-like accidents and provide its technical basis. • Point out that Fukushima accident might have been prevented if improved EOPs had been used. • Propose key points and suggestions for improving the EOPs. - Abstract: One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima

Depressurization-filtration system of the containment of French PWR's

International Nuclear Information System (INIS)

L'homme, A.; Schektman, N.

1987-01-01

In the hypothetical event of a core meltdown occurring in a pressurized water reactor, and in order to preserve the integrity of the containment threatened by a build-up in pressure, EDF has developed, with the CEA, a decompression device which filters the containment internal atmosphere by using an unused containment penetration, and a sand-box, as filtering mechanism. This device and its procedure for utilization, constitute the U5 procedure. Check-tests on a semi-industrial scale have been carried out at the Nuclear Research Centre at Cadarache, by using columns of sand 80 cm high, according to following varying criteria: the granulometry of the sand, that of the aerosols, the flow-through speed, and the percentage steam content of the fluid to be filtered. The filtering material chosen is sand of a median diameter of 0.6 mm. (log normal distribution). The purification factor is above 10. The device tested meets the chosen targets, and is applied today to French units on condition to simple modifications concerning specific aspects of different series. The first is expected to be put into service during 1987

ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

2000-03-01

A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

ROSA-II test data report, 10

International Nuclear Information System (INIS)

1977-12-01

Results of the ROSA-II test simulating a loss-of coolant accident (LOCA) in a light water reactor (LWR) are presented, including the test conditions and interpretation of the phenomena for test runs 415, 417, 421 and 422. Even in small break at the cold leg, the core is exposed to void and the temperature rises. In small break of the hot leg, however, core cooling keeps without temperature rise, because there still remains much residual water and upward core flow exists. Direct effect of the HPCI on the depressurization rate is small, but it increases the accumulator injection rate, leading to early core reflooding and early core cooling from upward. Effects of the secondary system depressurization are increase of depressurization and discharge rates of the primary loop, which results in early initiation of the accumulator injection and core reflooding. (auth.)

Probabilistic risk assessment (PRA) on the effectiveness of a core rescue system (SSN) for PWRs

International Nuclear Information System (INIS)

Petrangeli, G.; Valeri, A.

1983-01-01

Safety systems for the prevention of LWR core severe damage have recently been studied, which are based on automatic primary system depressurization and on borated water injection by low pressure accumulators. These systems have been named Core Rescue System (SSN). The present study evaluates the reduction in core melt probability brought about by the installation of a SSN system on the RSS (WASH 1400) PWR plant (Surry 1). The calculated result is a core melt probability reduction factor of about 250. Taking into account the possible effect of external or internal unknown events of negligible, yet undefined, probability it is concluded that a SSN system can make a plant ten times safer. The first part of a review report by Prof. N.C.Rasmussen, MIT, dealing with general comment, is attached

SFR Safety Consideration in Light of Fukushima Dai-ichi Accident

International Nuclear Information System (INIS)

Yamaguchi, Akira

2013-01-01

SFR Considerations: Fukushima Dai-ichi Accident: • Combined LORL and LOHS type initiated from SBO; • High pressure water-steam cooling system: – Depressurization - Not needed; – Ultimate heat sink - Robust (NC to atmosphere); – Continuous injection - Not needed (large sensible heat capacity). • Severe accident management: – RPV failure resulted in depressurization - Elevated temperature; – Heat sink to atmosphere - Freeing risk, sodium fire risk; – Mobile power supply - External resource may not be needed; – Seawater injection with fire engines - Sodium injection not needed; • Containment performance and accessibility: – Containment - Large containment volume and low pressure system; – Explosives - Sodium fire and hydrogen explosion

Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae K., E-mail: taekyu8@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Arcilesi, David J., E-mail: arcilesi.1@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Kim, In H., E-mail: ihkim0730@gmail.com [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Sun, Xiaodong, E-mail: sun.200@osu.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Christensen, Richard N., E-mail: rchristensen@uidaho.edu [Nuclear Engineering Program, The Ohio State University, Columbus, OH 43210 (United States); Oh, Chang H. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Kim, Eung S., E-mail: kes7741@snu.ac.kr [Idaho National Laboratory, Idaho Falls, ID 83402 (United States)

2016-04-15

Highlights: • Uncertainty quantification and benchmark study are performed to validate an ANSYS FLUENT computer model for a depressurization process in a high-temperature gas-cooled reactor. • An ANSYS FLUENT computer model of a 1/8th scaled-down geometry of a VHTR hot exit plenum is presented, which is similar to the experimental test facility that has been constructed at The Ohio State University. • Using the computer model of the scaled-down geometry, the effects of the depressurization process and flow oscillations on the subsequent density-driven stratified flow phenomenology are examined computationally. • The effects of the scaled-down hot exit plenum internal structure temperature on the density-driven stratified flow phenomenology are investigated numerically. - Abstract: An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment. Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study

Triaxial testing system for pressure core analysis using image processing technique

Science.gov (United States)

Yoneda, J.; Masui, A.; Tenma, N.; Nagao, J.

2013-11-01

In this study, a newly developed innovative triaxial testing system to investigate strength, deformation behavior, and/or permeability of gas hydrate bearing-sediments in deep sea is described. Transport of the pressure core from the storage chamber to the interior of the sealing sleeve of a triaxial cell without depressurization was achieved. An image processing technique was used to capture the motion and local deformation of a specimen in a transparent acrylic triaxial pressure cell and digital photographs were obtained at each strain level during the compression test. The material strength was successfully measured and the failure mode was evaluated under high confining and pore water pressures.

Risk impact of two accident management strategies

International Nuclear Information System (INIS)

Dingman, S.; Camp, A.

1992-01-01

This report probabilistic Risk Assessment is used to evaluate two accident management strategies: intentionally depressurizing the reactor coolant system of a pressurized water reactor to prevent containment-pressurization during high pressure melt ejection, and flooding the containment of a boiling water reactor to prevent or delay vessel breach. Sensitivity studies indicated that intentional depressurization would not provide a significant risk reduction at Surry. A preliminary evaluation of the containment flooding strategy indicated that it might prove beneficial for some plants, but that further strategy development would be needed to fully evaluate the strategy-

Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

International Nuclear Information System (INIS)

Park, Rae-Joon; Hong, Seong-Wan

2011-01-01

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

Thermosyphon Phenomenon as an alternate heat sink of Shutdown Cooling System for the CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jonghyun [GNEST, Seoul (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

2006-07-01

During the outage(overhaul) of the CANDU plant, there is a period when the coolant is partially drained to the reactor header level and the coolant is cooled and depressurized by Shutdown Cooling System(SDCS) other than PHTS pump. In the postulated accident of the loss of SDCS-the PHTS pump failure, the primary coolant system should be cooled by the alternate heat sink using the thermosyphon pheonomenon(TS) through the steam generator(SG) This study was aimed at verification and analyzing the core cooling ability of the TS. And the sensitivity analysis was done for the number of SGs used in the TS. As an analysis tool, RELAP5/CANDU was used.

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Laboratory Evaluation of Energy Recovery Ventilators

Energy Technology Data Exchange (ETDEWEB)

Kosar, D. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2016-11-01

Over the years, building scientists have characterized the relationship between building airtightness, exhaust-only appliances airflows, and building depressurization. Now, as the use of deep retrofit measures and new construction practices is growing to realize lower infiltration levels in increasingly tighter envelopes, performance issues can arise with the operation of exhaust-only appliances in a depressurized home. As the depressurization levels climb in tighter homes, many of these exhaust-only appliances see their rated airflows reduced and other related performance issues arise as a result. If sufficiently depressurized, atmospherically vented combustion appliances that may be present in the home can backdraft as well. Furthermore, when exhaust-only appliances operate and the tight home becomes depressurized, water vapor intrusion from outdoors can raise additional issues of mold in the building envelope in more humid climates.

Assessment of the potential for HPME during a station blackout in the Surry and Zion PWRS

International Nuclear Information System (INIS)

Knudson, D.L.; Bayless, P.D.; Dobbe, C.A.; Odar, F.

1994-01-01

The integrity of a PWR (pressurized water reactor) containment structure could be challenged by direct heating associated with a HPME (high pressure melt ejection) of core materials following reactor vessel lower head breach during certain severe accidents. Structural failure resulting from direct containment heating is a contributor to the risk of operating a PWR. Intentional RCS (reactor coolant system) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce those risks by mitigating the severity of the HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before failure of the lower head providing an unintentional mechanism for depressurization and HPME mitigation. This paper summarizes an assessment of RCS depressurization with respect to the potential for HPME during a station blackout in the Surry and Zion PWRs. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code and an evaluation of RCS depressurization-related probabilities primarily based on the code results

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)

Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

International Nuclear Information System (INIS)

Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

2013-01-01

SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

Development of a system code for transient analysis in a HTGR

International Nuclear Information System (INIS)

Lee, Tae Beom

2004-02-01

A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point

The assessment of RELAP5/MOD2 based on pressurizer transient experiments

International Nuclear Information System (INIS)

Xue Hanjun; Tanrikut, A.; Menzel, R.

1992-03-01

Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

International Nuclear Information System (INIS)

Park, R.J.; Kim, S.B.; Kim, H.D.

2004-01-01

High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

A preliminary study for the implementation of general accident management strategies

International Nuclear Information System (INIS)

Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung

1997-01-01

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated

PWR cold-leg small break loca with faulty HPI

International Nuclear Information System (INIS)

Kumamaru, H.; Kukita, Y.

1991-01-01

The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

Experiments on pollutant transport from soil into residential basements by pressure-driven airflow

International Nuclear Information System (INIS)

Nazaroff, W.W.; Lewis, S.R.; Doyle, S.M.; Moed, B.A.; Nero, A.V.

1987-01-01

At two residences in Portland, OR, they have investigated (1) the coupling between residential basements and the air in nearby soil and (2) the influence of basement depressurization on the migration of air in soil. With the basements depressurized 25-50 Pa relative to outdoor air, underpressures as great as 20-40% of those in the basement were observed at sampling points in the soil. Sulfur hexafluoride was injected into the soil near the houses and its concentration monitored in soil air and in the house over time, both with and without basement depressurization. Depressurization was seen to have a substantial effect on the migration of the tracer within the soil. For basement depressurizations of 25-50 Pa, effective transport velocities through the soil and into the houses were observed to exceed 1 m h -1 . Airborne 222 Rn concentration was monitored in the basement of one house during the 6-day investigation and was seen to increase substantially on each of the seven occasions that the house was depressurized. The techniques employed are applicable to the study of problems of excessive radon entry into buildings and the migration of toxic vapors from waste dumps and landfills

Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Hsu, C.J.

1983-01-01

In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

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

A preliminary study for the implementation of general accident management strategies

Energy Technology Data Exchange (ETDEWEB)

Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

A preliminary study for the implementation of general accident management strategies

Energy Technology Data Exchange (ETDEWEB)

Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

THEXSYST - a knowledge based system for the control and analysis of technical simulation calculations

International Nuclear Information System (INIS)

Burger, B.

1991-07-01

This system (THEXSYST) will be used for control, analysis and presentation of thermal hydraulic simulation calculations of light water reactors. THEXSYST is a modular system consisting of an expert shell with user interface, a data base, and a simulation program and uses techniques available in RSYST. A knowledge base, which was created to control the simulational calculation of pressurized water reactors, includes both the steady state calculation and the transient calculation in the domain of the depressurization, as a result of a small break loss of coolant accident. The methods developed are tested using a simulational calculation with RELAP5/Mod2. It will be seen that the application of knowledge base techniques may be a helpful tool to support existing solutions especially in graphical analysis. (orig./HP) [de

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

Passive afterheat removal in the HTGR with the liner cooling system as a heat sink

International Nuclear Information System (INIS)

Rehm, W.; Jahn, W.; Verfondern, K.

1984-09-01

The report deals with the transients of temperature and system pressure and the fission product behaviour in the primary circuit of an HTGR during passive afterheat removal, where the liner cooling system of the PCRV serves as a heat sink. The analysis has been made for the PNP-500-reactor representing nuclear plants with medium thermal power. The investigations show that the liner cooling system is able to control a core heatup. High temperature loads are encountered in the upper core region. In the case of a reactor under pressure the fuel elements and the primary circuit remain intact as the first and second barriers for fission products. In the case of a depressurized primary circuit the liner cooling system also keeps the PCRV at normal operating temperatures. The effects of a core heatup on component damage and release of fission products are thus limited. (orig.) [de

Leak detection system design and operating considerations for the US-CRBRP

International Nuclear Information System (INIS)

Kruger, G.B.; Eng, K.Y.; Kelly, W.L.

1976-01-01

Diffusion membrane type hydrogen detectors are provided for monitoring the sodium exiting each evaporator and superheater in the Clinch River Breeder Reactor Plant. These detectors allow detection of small water to sodium leaks and provide the plant operator with an early warning signal. Hydrogen detectors are located at the exit sodium streams of each steam generator module, the vent from the module semi-stagnant region, the cold leg piping, and in an intermediate system sodium expansion tank cover gas region. In addition, an electrochemical oxygen detector is located in the cold leg piping. The leak detection system is capable of detecting the presence of steam/water leaks on the order of 0.45 x 10 -5 kg/sec or larger and of signaling within one to three minutes upon initiation of a leak, during normal operation. Operator action is taken upon receipt of a leak signal to shutdown the affected system, by closing steam/water isolation valves and depressurizing the affected unit

Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

Directory of Open Access Journals (Sweden)

Andrej Prošek

2013-01-01

Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

Depressurization of a spread of Brazil-Bolivia gas pipeline and the emergency repair of a weld crack in a instrument derivation at Campo Grande compression station; Despressurizacao de trecho do gasoduto Bolivia-Brazil para reparo emergencial de trinca em uma derivacao de instrumento de temperatura na Estacao de Compressao de Campo Grande - MS

Energy Technology Data Exchange (ETDEWEB)

Dietrich, Carlos Ribeiro; Leite Junior, Ismael Casano [TBG - Transportadora Brasileira Gasoduto Bolivia Brasil S.A., Rio de Janeiro, RJ (Brazil)

2005-07-01

The purpose of this paper is to report the actions taken to repair a gas leak, at an original pressure of 100 kgf/cm{sup 2}, occurred due to a 1 1/2'' branch pipe weld crack, located on the 24'' Campo Grande - Mato Grosso do Sul Compression Station discharge pipe. This branch pipe was used to a thermo well installation and was submitted to an additional strength caused by thermo well vibration. The weld repair actions required an urgent depressurization of a 33 km spread of Bolivia-Brazil Pipeline in a timely manner, to avoid any negative impact in the operational schedule. (author)

Irradiated ignition over solid materials in reduce pressure environment: Fire safety issue in man-made enclosure system

Science.gov (United States)

Nakamura, N.; Aoki, A.

Effects of ambient pressure and oxygen yield on irradiated ignition characteristics over solid combustibles have been studied experimentally Aim of the present study is to elucidate the flammability and chance of fire in depressurized enclosure system and give ideas for the fire safety and fire fighting strategies in such environment Thin cellulosic paper is considered as the solid combustible since cellulose is one of major organic compounds and flammables in the nature Applied atmosphere consists of inert gas either CO2 or N2 and oxygen and various mixture ratios are of concerned Total ambient pressure level is varied from 0 1MPa standard atmospheric pressure to 0 02MPa Ignition is initiated by external thermal flux exposed into the solid surface as a model of unexpected thermal input to initiate the localized fire Thermal degradation of the solid induces combustible gaseous products e g CO H2 or other low class of HCs and the gas mixes with ambient oxygen to form the combustible mixture over the solid Heat transfer from the hot irradiated surface into the mixture accelerates the local exothermic reaction in the gas phase and finally thermal runaway ignition is achieved Ignition event is recorded by high-speed digital video camera to analyze the ignition characteristics Flammable map in partial pressure of oxygen Pox and total ambient pressure Pt plane is made to reveal the fire hazard in depressurized environment Results show that wider flammable range is obtained depending on the imposed ambient

Engineering reliability in design phase: An application to AP-600 reactor passive safety system

International Nuclear Information System (INIS)

Majumdr, D.; Siahpush, A.S.; Hills, S.W.

1992-01-01

A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

Radiation effluent suppression system

International Nuclear Information System (INIS)

Watanabe, Atsushi.

1992-01-01

In a radiation release suppression system upon accident, an electromotive valve, a pneumatic operation valve or a manual operation valve is disposed to gas ventilation pipelines which are extended from both of a dry well and a wet well of a reactor container to a stuck. In addition, a combination filter of a metal fiber filter made of stainless steel etc. and an activated carbon fiber filter is disposed in the midway of pipelines in a reactor building. With such a constitution, the inside of the container can be depressurized (prevention of ruptures) and the amount of radioactive substances released to circumstances is remarkably suppressed by the effect of radioactive substance capturing effect of the metal fiber filter made of stainless steel etc. disposed in the vent pipe in the container and a radioactive substance capturing effect by the combination filter of the metal fiber filter made of stainless steel, etc. and the activated carbon fiber filter disposed in the gas ventilation pipelines even upon occurrence of an accident exceeding design basis. Systems can be simplified and minimized, and cost down can also be attained. (N.H.)

Development and design of a high pressure carbon dioxide system for the separation of hazardous contaminants from non-hazardous debris

International Nuclear Information System (INIS)

Adkins, C.L.J.; Russick, E.M.; Smith, H.M.; Olson, R.B.

1995-01-01

Under the Department of Energy (DOE)/United States Air Force (USAF) Memorandum of Understanding, a system is being designed that will use high pressure carbon dioxide for the separation of oils, greases, and solvents from non-hazardous solid waste. The contaminants are dissolved into the high pressure carbon dioxide and precipitated out upon depressurization. The carbon dioxide solvent can then be recycled for continued use. Excellent extraction capability for common manufacturing oils, greases, and solvents has been measured. It has been observed that extraction performance follows the dilution model if a constant flow system is used. The solvents tested are extremely soluble and have been extracted to 100% under both liquid and mild supercritical carbon dioxide conditions. These data are being used to design a 200 liter extraction system

Assessment of thermalhydraulic phenomena for external water make-up

International Nuclear Information System (INIS)

Harwood, C.; Baschuk, J.

2015-01-01

Following the Fukushima Daiichi accident, Canadian NPP licensees implemented a number of changes, including additional provisions for water make-up to the reactor/boiler systems. The CNSC has placed a contract with CNL to model some of the make-up options, focusing on cooling via the boilers to prevent core damage. Such strategies have been credited with sustaining thermo syphoning in the primary system and thus prolonging the available time for the operator to provide pumped make-up to the boilers or emergency coolant injection to the core, thereby maintaining decay heat removal. This paper presents results of CATHENA calculations of an extended loss of all electrical power in which the operator manually depressurizes the boilers by crash-cooling, thus allowing water to flow to the boilers by gravity from the deaerator tank. The rapid cooling of the boilers promotes thermo-syphoning flow in the primary heat transport system and results in a corresponding cool-down and depressurization of this system. (author)

Assessment of thermalhydraulic phenomena for external water make-up

Energy Technology Data Exchange (ETDEWEB)

Harwood, C., E-mail: Christopher.Harwood@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, ON (Canada); Baschuk, J. [Canadian Nuclear Laboratories, Chalk River, ON, (Canada)

2015-07-01

Following the Fukushima Daiichi accident, Canadian NPP licensees implemented a number of changes, including additional provisions for water make-up to the reactor/boiler systems. The CNSC has placed a contract with CNL to model some of the make-up options, focusing on cooling via the boilers to prevent core damage. Such strategies have been credited with sustaining thermo syphoning in the primary system and thus prolonging the available time for the operator to provide pumped make-up to the boilers or emergency coolant injection to the core, thereby maintaining decay heat removal. This paper presents results of CATHENA calculations of an extended loss of all electrical power in which the operator manually depressurizes the boilers by crash-cooling, thus allowing water to flow to the boilers by gravity from the deaerator tank. The rapid cooling of the boilers promotes thermo-syphoning flow in the primary heat transport system and results in a corresponding cool-down and depressurization of this system. (author)

Static and transient characteristics of the shaft seal system for helium gas circulator (Part 1)

International Nuclear Information System (INIS)

Morohoshi, S.; Saki, K.; Nemoto, M.; Taniguchi, S.; Sugimoto, A.; Kojima, M.

1982-01-01

A development program of the shaft seal system for the helium circulator supported by water lubricated bearings is presented. A seal system simulating tester and a computer program which can simulate the transient characteristics of a buffer gas seal system were newly introduced, and an investigation was performed experimentally and analytically of the characteristics of water and gas seals and of the buffer gas seal system including the control system. Main results are as follows: (1) Water seals were especially investigated in detail, and it was found that turbulence in water flow through seal clearance and deformation of seal components affected the leakage characteristics of water seals. They should be considered not only to make safety design but also to get optimum design of the seal system. (2) The calculation method for transient response of the buffer gas seal system including the control system was developed. This digital simulating method can well simulate transients encountered in the tester, and it would make a powerful tool for developing a safe seal system under steady state operation conditions and at depressurization accidents in a reactor

Exchange of pressurizer safeguarding system at Biblis nuclear power station

International Nuclear Information System (INIS)

Weber, D.; Hofbeck, W.

1991-01-01

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

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

76 FR 10482 - Special Conditions: Boeing Model 787-8 Airplane; Overhead Flightcrew-Rest Compartment Occupiable...

Science.gov (United States)

2011-02-25

... out. (b) For a manual, hand-held extinguishing system (designed as the sole means to fight a fire or..., except that Sec. 25.1309 remains at Amendment 25- 117 for cargo-fire protection systems. If the... to two flight crewmembers trained in the evacuation, fire fighting, and depressurization procedures...

Gest-sip1 experiments and post-test calculations with the relap5 code

International Nuclear Information System (INIS)

Achilli, A.; Cattadori, G.; Ferri, R.; Gandolfi, S.; Bianchi, F.; Meloni, P.

2001-01-01

The SIP-1 apparatus (Sistema di Iniezione Passiva) was conceived, designed, numerically simulated and tested by the SIET company as an innovative depressurization and make-up device for the New Generation LWRs. In particular it is suitable to cope with those accidents where pressure in the circuit must be dumped to allow low pressure injection systems to intervene. The main peculiarity of SIP-1 is the capability of de-pressurizing a system by cold water injection, rather than by discharging mass to the outlet, as in the common depressurization systems. ENEA sponsored all the research activity, starting from the SIP-1 design, its numerical simulation with the Relap5 code, the realisation of an experimental facility up to the test execution and post-test calculations. An experimental campaign on the GEST-SIP1 facility was performed in July 2000. The facility is mainly constituted by a U-tube Steam Generator which a proper model of SIP-1 apparatus is connected to. A series of Small Break LOCAs was simulated by varying the break size and different steady conditions were investigated to verify the stability of SIP-1, the lack of unexpected interventions and the actuation modalities. This paper deals with the description of the GEST-SIP1 experimental facility, the SIP-1 operating principles, the most meaningful results of the tests and the capability of the Relap5 code in reproducing phenomena and events. (author)

Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

Energy Technology Data Exchange (ETDEWEB)

Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-03-16

During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

International Nuclear Information System (INIS)

Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

2005-01-01

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

International Nuclear Information System (INIS)

Lap-Yan, C.; Wie, T. Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

Assessment of the potential for high-pressure melt ejection resulting from a Surry station blackout transient

International Nuclear Information System (INIS)

Knudson, D.L.; Dobbe, C.A.

1993-11-01

Containment integrity could be challenged by direct heating associated with a high pressure melt ejection (HPME) of core materials following reactor vessel breach during certain severe accidents. Intentional reactor coolant system (RCS) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce risks by mitigating the severity of HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before vessel breach, providing an alternate mechanism for RCS depressurization and HPME mitigation. This report contains an assessment of the potential for HPME during a Surry station blackout transient without operator action and without recovery. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code to calculate the plant response with and without hot leg countercurrent natural circulation, with and without reactor coolant pump seal leakage, and with variations on selected core damage progression parameters. RCS depressurization-related probabilities were also evaluated, primarily based on the code results

Testing of indoor radon-reduction techniques in basement houses having adjoining wings. Final report, August 1988-September 1989

International Nuclear Information System (INIS)

Messing, M.

1990-11-01

The report gives results of tests of indoor radon reduction techniques in 12 existing Maryland houses, with the objective of determining when basement houses with adjoining wings require active soil depressurization (ASD) treatment of both wings, and when treatment of the basement alone is sufficient. In five basement houses with adjoining slabs on grade, ASD treatment of both wings provided an incremental additional radon reduction of 0 to 5.2 pCi/L, compared to ASD treatment of either one of the slabs alone. However, basement-only treatment reduced radon to <4 pCi/L in all five houses. In six basement houses having adjoining crawl spaces, ASD treatment of both wings (including sub-liner depressurization of the crawl space) provided little additional reduction compared to basement-only treatment, when sub-slab communication was good. When communication was not good, treatment of both wings was required to achieve <4 pCi/L. Tests of one fully slab-on-grade house showed that, when there is good aggregate under the slab, a one-pipe sub-slab depressurization system can achieve <1-2 pCi/L, even when there are forced-air supply ducts under the slab

Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

International Nuclear Information System (INIS)

Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

1996-01-01

Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits

Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

International Nuclear Information System (INIS)

Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

2011-01-01

Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

Steam line break analysis in CAREM-25 reactor

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

2000-01-01

The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

Steam Line Break Analysis in CAREM-25 Reactor

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo; Vertullo, Alicia; Garcia, A; Schlamp, Miguel

2000-01-01

The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model.The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator.As a consequence and due to reactor features the core power is also increased.As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided.Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System.In all the sequences the DNBR and CPR remain above the minimum safety values established by design.Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised

Prednisolone multicomponent nanoparticle preparation by aerosol solvent extraction system.

Science.gov (United States)

Moribe, Kunikazu; Fukino, Mika; Tozuka, Yuichi; Higashi, Kenjirou; Yamamoto, Keiji

2009-10-01

Prednisolone nanoparticles were prepared in the presence of a hydrophilic polymer and a surfactant by the aerosol solvent extraction system (ASES). A ternary mixture of prednisolone, polyethylene glycol (PEG), and sodium dodecyl sulfate (SDS) dissolved in methanol was sprayed through a nozzle into the reaction vessel filled with supercritical carbon dioxide. After the ASES process was repeated, precipitates of the ternary components were obtained by depressurizing the reaction vessel. When a methanolic solution of prednisolone/PEG 4000/SDS at a weight ratio of 1:6:2 was sprayed under the optimized ASES conditions, the mean particle size of prednisolone obtained after dispersing the precipitates in water was observed to be ca. 230 nm. Prednisolone nanoparticles were not obtained by the binary ASES process for prednisolone, in the presence of either PEG or SDS. Furthermore, ternary cryogenic cogrinding, as well as solvent evaporation, was not effective for the preparation of prednisolone nanoparticles. As the ASES process can be conducted under moderate temperature conditions, the ASES process that was applied to the ternary system appeared to be one of the most promising methods for the preparation of drug nanoparticles using the multicomponent system.

PHYSICAL AND NUMERICAL MODELING OF ASD EXHAUST DISPERSION AROUND HOUSES

Science.gov (United States)

The report discusses the use of a wind tunnel to physically model the dispersion of exhaust plumes from active soil depressurization (ASD) radon mitigation systems in houses. he testing studied the effects of exhaust location (grade level vs. above the eave), as house height, roo...

LOCA analysis of the IRIS reactor

International Nuclear Information System (INIS)

Bajs, T.; Grgic, D.; Cavlina, N.

2003-01-01

The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

Performance assessment of containment filtered venting system with Venturi scrubber

International Nuclear Information System (INIS)

Adinarayna, K.N.V.; Ali, Seik Mansoor; Balasubramaniyan, V.

2015-01-01

Venting through appropriate filtration systems is now being considered as a severe accident management strategy for maintaining the containment integrity and also as a means to reduce the radiological consequences to the public and environment. The option of filtered containment venting appears to have assumed significance in the post- Fukushima accident backdrop. Back-fitting of a suitable Venturi scrubber based CFVS for the Indian BWRs (TAPS- 1 and 2) at Tarapur is now being contemplated. Several key issues need to be carefully addressed for ensuring the desired functional capability of such a system. At the outset, this paper highlights a few thermal hydraulic issues that are of interest from regulatory perspective. This is followed by a detailed description of the mathematical models developed for assessing the depressurization characteristics of CFVS, energy absorption capacity of the Scrubber Tank (ST) water inventory, iodine removal and aerosol retention capability etc. Finally, application of these models to investigate the response of CFVS under twin unit SBO conditions in TAPS-1 and 2 is presented. The studies presented here give insight into the key variables affecting the CFVS performance and would be useful to both the system designer as well as the regulator. (author)

Natural circulation cooldown analysis for Yonggwang 3 and 4 per US NRC BTP RSB 5-1 requirements

International Nuclear Information System (INIS)

Seo, J.T.; Ko, C.S.; Ro, T.S.; Simoni, L.P.

2004-01-01

The Natural Circulation Cooldown (NCC) analysis from normal operations to shutdown cooling entry conditions for Yonggwang units 3 and 4 (YGN 3 and 4) was performed within the requirements of U.S. Nuclear Regulatory Commission (NRC) Branch Technical Position (BTP) RSB 5-1. The results showed that the YGN 3 and 4 can be cooled and depressurized to the shutdown entry conditions (350 deg F, 410 psia) within 16 hours under natural circulation condition requiring only 78% of the minimum condensate water storage capacity in conformance with BTP RSB 5-1 requirements. The results also demonstrated that the safety grade Reactor Coolant Gas Vent System (RCGVS) has sufficient capacity for the RCS depressurization as well as for the steam void control in the reactor vessel upper head region. (author)

Methodology on the sparger development for Korean next generation reactor

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K.

1999-06-01

In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloud acceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs

Methodology on the sparger development for Korean next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K

1999-06-01

In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloudacceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs.

Thermal hydraulic analysis of BWR containment venting system

International Nuclear Information System (INIS)

Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

2015-01-01

Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

Experiment data report for semiscale Mod-1 test S-02-3 (blowdown heat transfer test)

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1975-09-01

Recorded test data are presented for Test S-02-3 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-3 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,263 psig. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with a moderately heated core (75 percent design power level). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was also set at the 75 percent design level to achieve full core temperature differential. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.2 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core coolant injection or pressure suppression system coolant spray

Simulation of a postulated 2% cold leg break in Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Palmieiri, Elcio Tadeu; Azevedo, Carlos Vicente Goulart de; Aronne, Ivan Dionysio

2007-01-01

This paper presents the simulation of a 2% break in the cold leg pipe of Angra 2 nuclear power plant, with the computer code RELAP5/Mod3.3. The main boundary conditions specified for this simulation were: no injection from high pressure injection system; enhanced depressurization of the primary system by opening the pressure operated relief valve (PORV) and the safety relief valve (SRV) when core temperature reaches circa 100 K above saturation; and accumulator injection starting at 2.7 MPa. The specific objectives to be addressed with this simulation are: the core boil-off and dryout at relatively high pressure in the primary system; the phenomena during enhanced primary depressurization; the effectiveness of hot leg accumulator injection into the partially uncovered rod bundle; and the core rewetting. The results obtained were compared with the Lobi A1-93 test, which was performed under the same boundary conditions. This activity was executed in the scope of IAEA research project Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3 Code Applying CIAU Methodology (author)

A Pilot Study on Applying Risk Informed Application Option 2 to Six Systems in UCN 3

International Nuclear Information System (INIS)

Kim, Kil-Yoo; Yang, Joon-Eon; Lee, Young-Joo; Chung, Hye-Won

2007-01-01

To reduce the unnecessary burden of a regulation, NRC prepared three options for the risk informed regulatory framework known as Option 1, Option 2 and Option 3. In Option 2, all safety related Structure, System and Components (SSCs) and non-safety related SSCs are evaluated from a safety point of view, and the low safety significant SSCs belonging to the safety related group are called 'Risk Informed Safety Class (RISC) - 3' SSCs. The 'RISC-3' SSCs can be exempted from the special treatment requirements such as a seismic and environmental requirement, of 10 CFR 50. Two years ago, a paper was published which described the Option 2 method applied to the high pressure safety injection system (HPSI) and the essential service water system (ESW) of UCN 3. However, this paper describes the results when Option 2 is applied to the other 4 systems such as a low pressure safety injection system(LPSI), safety depressurization system(SDS), instrument air system(IAS), safety injection tank(SIT). First of all, this paper includes the results from the importance analysis in view of a Fire PSA and Level 2 PSA

A gas production system from methane hydrate layers by hot water injection and BHP control with radial horizontal wells

Energy Technology Data Exchange (ETDEWEB)

Yamakawa, T.; Ono, S.; Iwamoto, A.; Sugai, Y.; Sasaki, K. [Kyushu Univ., Fukuoka, Fukuoka (Japan)

2010-07-01

Reservoir characterization of methane hydrate (MH) bearing turbidite channel in the eastern Nankai Trough, in Japan has been performed to develop a gas production strategy. This paper proposed a gas production system from methane hydrate (MH) sediment layers by combining the hot water injection method and bottom hole pressure control at the production well using radial horizontal wells. Numerical simulations of the cylindrical homogeneous MH layer model were performed in order to evaluate gas production characteristics by the depressurization method with bottom hole pressure control. In addition, the effects of numerical block modeling and averaging physical properties of MH layers were presented. According to numerical simulations, combining the existing production system with hot water injection and bottom hole pressure control results in an outward expansion of the hot water chamber from the center of the MH layer with continuous gas production. 10 refs., 15 figs.

20% inlet header break analysis of Advanced Heavy Water Reactor

International Nuclear Information System (INIS)

Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

2001-01-01

The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

Mathematical Model and Simulation of Gas Hydrate Reservoir Decomposition by Depressurization Modèle mathématique et simulation de dépressurisation et de décompression d’un réservoir d’hydrates de méthane

Directory of Open Access Journals (Sweden)

Zhao J.

2012-05-01

Full Text Available The numerical model for the depressurization of methane hydrates in a confined reservoir is presented based on mass conservation in porous media, incorporating multiphase flow theory and kinetics of gas hydrate dissociation. The universal implicit difference method is adopted, and the corresponding computer program is developed. During the production of the hydrate reservoir, distribution and the physical changes are analyzed and the gas hydrate dissociation and gas production law are studied from the computation. A numerical simulation shows that the reservoir pressure is descending slowly, which benefits the stabilization of the reservoir and inevitably decreases the efficiency in the production of gas hydrates in the depressurizing process. The gas production rate is controlled by the well pressure. The results are presented to show how this model may be used to estimate a lower downhole pressure of the well for hydrate recovery and how these results depend on reservoir and hydrate properties. Le modèle numérique présenté ici simule la dépressurisation d’hydrates de méthane dans un réservoir confiné; il se base sur le principe de conservation de la masse en milieu poreux, en intégrant la théorie de l’écoulement polyphasique et la cinétique de dissociation des hydrates de méthane. La méthode implicite et universelle des différences finies est utilisée et le programme informatique qui s’y rapporte est développé. Lors de l’exploitation du réservoir d’hydrates de méthane, la répartition et les changements physiques sont analysés et les lois sur la dissociation des hydrates de méthane et la production de gaz sont étudiées à partir des calculs. Une simulation numérique montre que la pression dans le réservoir diminue lentement, ce qui permet au réservoir de se stabiliser et diminue inévitablement le rendement de l’exploitation d’hydrates de méthane lors du processus de dépressurisation. Le rythme de

Development and industrial tests of the first LNG hydraulic turbine system in China

OpenAIRE

Jie Chen; Yihuai Hua; Qingbo Su; Xueli Wan; Zhenlin Li

2016-01-01

The cryogenic hydraulic turbine can be used to replace the conventional J–T valve for LNG or mixed refrigerant throttling and depressurization in a natural gas liquefaction plant. This advanced technology is not only to enhance the efficiency of the liquefaction plant, but to usher a new trend in the development of global liquefaction technologies. China has over 136 liquefaction plants, but the cryogenic hydraulic turbines have not been deployed in industrial utilization. In addition, these ...

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

Science.gov (United States)

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

2018-01-01

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

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

Directory of Open Access Journals (Sweden)

Clamens Olivier

2018-01-01

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

Sediment–well interaction during depressurization

KAUST Repository

Shin, Hosung; Santamarina, Carlos

2016-01-01

production from hydrate accumulations in marine sediments. Sediment–well interaction is examined using a nonlinear finite element simulator. The hydro-mechanically coupled model represents the sediment as a Cam-Clay material, uses a continuous function

Iodine removal in containment filtered venting system during nuclear accident

International Nuclear Information System (INIS)

Bera, Subrata; Deo, Anuj Kumar; Nagrale, D.B.; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

2015-01-01

Post Fukushima nuclear accident, containment filtered venting system is being introduced in Indian nuclear power plant to strengthen the defense in depth safety barrier by depressurizing the containment building along with minimization of radioactivity release to environment during a severe accident. Radioactive iodine is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the most efficient design of CFVS, wet scrubbing mechanism has been employed through use of venture scrubber. The Iodine removal process in wet scrubber involves two processes: chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper, venturi has been modeled using the Calvert model. The variation of efficiency has been estimated for the different particle sizes. The impact of the shape parameter of log-normal distribution on the amount of scrubbed iodine has also been assessed. Release phase wise the scrubbed amount of iodine in the venturi based CFVS system has been estimated for a typical BWR. (author)

Risk factors of methane hydrate resource development in the concentrated zones distributed in the eastern Nankai Trough

Science.gov (United States)

Yamamoto, K.; Nagakubo, S.

2009-04-01

Some environmental and safety concerns on the offshore methane hydrate development have been raised, but the ground of such allegations are sometime not fully reasonable. The risks of methane hydrate resource development to environment and safety should be discussed upon methane hydrate occurrences condition, the production methods, and the designs of production system, under comprehensively scientific manners. In the Phase 1 of the Methane Hydrate Exploitation Program in Japan (FY2001-2008), the Research Consortium for Methane Hydrate Resources in Japan (MH21 Research Consortium) found methane hydrate concentrated zones in the eastern Nankai Trough that are potential prospects for resource development. The concentrated zones are consisted of turbidite-derived sandy sediments and hydrate crystals in pore spaces of sand grains (pore-filling type structure). The MH21 Research Consortium proposed the depressurization method as prime technique due to its efficiency of gas production in such concentrated zones, and has tried to develop conceptual designs of production systems based on the information of existing devices and facilities. Under the condition and circumstances described above, the authors tried to extract and evaluate some risk factors concerning methane hydrate development using depressurization in the area. Leakage of methane gas, that is less harmful substance to ecosystem than heavier hydrocarbons, from production system can be one possible risk. However, in the case of gas production through wellbore, even if catastrophic damages happen in the subsea production system during gas production, the leakages do not continue because the borehole could be filled by seawater and depressurization is stopped immediately. Another possible risk is a leakage of produced gas through seafloor. If methane hydrate production makes high pressure or temperature zones in sediments, the risk should be considered. However, depressurization method makes opposite condition

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.

Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

International Nuclear Information System (INIS)

Ruggles, A.E.; Cheng, L.Y.; Dimenna, R.A.; Griffith, P.; Wilson, G.E.

1994-06-01

A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis

BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

1985-03-01

This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

Pressure vessel SBLOCA simulation with trace: application to ISTF (Rosa V) - 151

International Nuclear Information System (INIS)

Abella, V.; Gallardo, S.; Verdu, G.

2010-01-01

In this work, an overview of the results obtained in the simulation of an Upper Head Small Break Loss-Of-Coolant-Accident (SBLOCA) under the assumption of total failure of High Pressure Injection System (HPIS) in the Large Scale Test Facility (LSTF) is provided. In previous works, an SBLOCA located in the Pressure Vessel (PV) Lower Plenum was simulated with TRACE. In that case, an asymmetrical steam generator secondary-side depressurization was produced as an accident management action at the Steam Generator in loop without pressurizer after the generation of safety injection signal to achieve a determined depressurization rate in the primary system. The new SBLOCA scenario has been simulated and results compared with experimental values, with the purpose of completing the analysis of PV SBLOCA. This study is developed in the frame of the OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA). Finally, the present paper represents a contribution for the study of safety analysis of vessel SBLOCAs and the assessment of the predictability of thermal-hydraulic codes like TRACE. (authors)

Research of working pulsation in closed angle based on rotating-sleeve distributing-flow system

Science.gov (United States)

Zhang, Yanjun; Zhang, Hongxin; Zhao, Qinghai; Jiang, Xiaotian; Cheng, Qianchang

2017-08-01

In order to reduce negative effects including hydraulic impact, noise and mechanical vibration, compression and expansion of piston pump in closed volume are used to optimize the angle between valve port and chamber. In addition, the mathematical model about pressurization and depressurization in pump chamber are analyzed based on distributing-flow characteristic, and it is necessary to use simulation software Fluent to simulate the distributing-flow fluid model so as to select the most suitable closed angle. As a result, when compression angle is 3°, the angle is closest to theoretical analysis and has the minimum influence on flow and pump pressure characteristic. Meanwhile, cavitation phenomenon appears in pump chamber in different closed angle on different degrees. Besides the flow pulsation is increasingly smaller with increasing expansion angle. Thus when expansion angle is 2°, the angle is more suitable for distributing-flow system.

Identification of advective entry of soil-gas radon into a crawl space covered with sheets of polyethylene foil

International Nuclear Information System (INIS)

Andersen, C.; Koopmanns, M.; Meijer, R.J. de

1996-04-01

To assess the effectiveness of mitigative measures against radon ( 222 Rn) entry into houses, experiments were conducted in a crawl-space house where the dirt floor of the crawl space was covered with sheets of 0.23 mm polyethylene foil fixed to the walls. The radon concentration was measured below the foil and in the crawl space together with environmental variables such as indoor-outdoor pressure differences. The experimental data was analyzed using various types of models including a simplistic mass-balance model, a regression model, and a two-dimensional numerical model based on Darcy flow or soil gas and combined diffusive and advective transport of radon. The main outcome of the work was that: (i) The soil-gas entry rate per pascal depressurization was at the order of 1 m 3 h -1 , (ii) the stack-related part of the depressurization of the crawl space (approx. 0.1 Pa deg. C -1 ) was controlled by the temperature difference between the living room of the house and the outdoors (not by the difference between the crawl space and the outdoors), (iii) that part of the wind-related depressurization that was measured by the pressure transducers seemed to force radon into the crawl space in the same proportion as the stack-related part of the depressurization, (iv) the ratio of advective and diffusive entry was approx. 0.7, when the crawl space was depressurized 1.5 Pa, (v) the effective diffusivity of the foil was found to be three orders of magnitude larger than that measured in the laboratory (the enhanced diffusivity was most likely caused by leaks in the foil and by mixing fans located in the crawl space), and (vi) there was no measurable mitigative impact of having the sheets of foil on the crawl-space floor even if the crawl space was artificially pressurized or depressurized. (au) 28 tabs., 36 ills., 61 refs

Identification of advective entry of soil-gas radon into a crawl space covered with sheets of polyethylene foil

Energy Technology Data Exchange (ETDEWEB)

Andersen, C. [Risoe National Lab., Dept. of Nucl. Safety Res. and Nucl. Facilities, Roskilde (Denmark); Koopmanns, M.; Meijer, R.J. de [Kernfysische Versneller Inst., Environmental Radioactivity Res., Groningen (Netherlands)

1996-04-01

To assess the effectiveness of mitigative measures against radon ({sup 222}Rn) entry into houses, experiments were conducted in a crawl-space house where the dirt floor of the crawl space was covered with sheets of 0.23 mm polyethylene foil fixed to the walls. The radon concentration was measured below the foil and in the crawl space together with environmental variables such as indoor-outdoor pressure differences. The experimental data was analyzed using various types of models including a simplistic mass-balance model, a regression model, and a two-dimensional numerical model based on Darcy flow or soil gas and combined diffusive and advective transport of radon. The main outcome of the work was that: (i) The soil-gas entry rate per pascal depressurization was at the order of 1 m{sup 3} h{sup -1}, (ii) the stack-related part of the depressurization of the crawl space (approx. 0.1 Pa deg. C{sup -1}) was controlled by the temperature difference between the living room of the house and the outdoors (not by the difference between the crawl space and the outdoors), (iii) that part of the wind-related depressurization that was measured by the pressure transducers seemed to force radon into the crawl space in the same proportion as the stack-related part of the depressurization, (iv) the ratio of advective and diffusive entry was approx. 0.7, when the crawl space was depressurized 1.5 Pa, (v) the effective diffusivity of the foil was found to be three orders of magnitude larger than that measured in the laboratory (the enhanced diffusivity was most likely caused by leaks in the foil and by mixing fans located in the crawl space), and (vi) there was no measurable mitigative impact of having the sheets of foil on the crawl-space floor even if the crawl space was artificially pressurized or depressurized. (au) 28 tabs., 36 ills., 61 refs.

Containment heat removal system

International Nuclear Information System (INIS)

Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

1992-01-01

This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

Solar thermal power system

Science.gov (United States)

Bennett, Charles L.

2010-06-15

A solar thermal power generator includes an inclined elongated boiler tube positioned in the focus of a solar concentrator for generating steam from water. The boiler tube is connected at one end to receive water from a pressure vessel as well as connected at an opposite end to return steam back to the vessel in a fluidic circuit arrangement that stores energy in the form of heated water in the pressure vessel. An expander, condenser, and reservoir are also connected in series to respectively produce work using the steam passed either directly (above a water line in the vessel) or indirectly (below a water line in the vessel) through the pressure vessel, condense the expanded steam, and collect the condensed water. The reservoir also supplies the collected water back to the pressure vessel at the end of a diurnal cycle when the vessel is sufficiently depressurized, so that the system is reset to repeat the cycle the following day. The circuital arrangement of the boiler tube and the pressure vessel operates to dampen flow instabilities in the boiler tube, damp out the effects of solar transients, and provide thermal energy storage which enables time shifting of power generation to better align with the higher demand for energy during peak energy usage periods.

Solution of multiple circuits of steam cycle HTR system

International Nuclear Information System (INIS)

Li, Fu; Wang, Dengying; Hao, Chen; Zheng, Yanhua

2014-01-01

In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated

Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

2015-07-15

Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

Experiment data report for Semiscale Mod-1 test S-02-5 (blowdown heat transfer test)

International Nuclear Information System (INIS)

1975-12-01

Recorded test data are presented for Test S-02-5 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-5 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 water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-5 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,253 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66 0 F. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling occurs

Numerical Simulation of the Depressurization Process of a Natural Gas Hydrate Reservoir: An Attempt at Optimization of Field Operational Factors with Multiple Wells in a Real 3D Geological Model

Directory of Open Access Journals (Sweden)

Zhixue Sun

2016-09-01

Full Text Available Natural gas hydrates, crystalline solids whose gas molecules are so compressed that they are denser than a typical fluid hydrocarbon, have extensive applications in the areas of climate change and the energy crisis. The hydrate deposit located in the Shenhu Area on the continental slope of the South China Sea is regarded as the most promising target for gas hydrate exploration in China. Samples taken at drilling site SH2 have indicated a high abundance of methane hydrate reserves in clay sediments. In the last few decades, with its relatively low energy cost, the depressurization gas recovery method has been generally regarded as technically feasible and the most promising one. For the purpose of a better acquaintance with the feasible field operational factors and processes which control the production behavior of a real 3D geological CH4-hydrate deposit, it is urgent to figure out the effects of the parameters such as well type, well spacing, bottom hole pressure, and perforation intervals on methane recovery. One years’ numerical simulation results show that under the condition of 3000 kPa constant bottom hole pressure, 1000 m well spacing, perforation in higher intervals and with one horizontal well, the daily peak gas rate can reach 4325.02 m3 and the cumulative gas volume is 1.291 × 106 m3. What’s more, some new knowledge and its explanation of the curve tendency and evolution for the production process are provided. Technically, one factor at a time design (OFAT and an orthogonal design were used in the simulation to investigate which factors dominate the productivity ability and which is the most sensitive one. The results indicated that the order of effects of the factors on gas yield was perforation interval > bottom hole pressure > well spacing.

Gas-cooled fast reactor safety

International Nuclear Information System (INIS)

Rickard, C.L.; Simon, R.H.; Buttemer, D.R.

1977-01-01

Initial conceptual design work on the GCFR began in the USA in the early 1960s and since the later 1960s has proceeded with considerable international cooperation. A 300 MWe GCFR demonstration plant employing three main cooling loops is currently being developed at General Atomic. A major preapplication licensing review of this demonstration plant was initiated in 1971 leading in 1974 to publication of a Safety Evaluation Report by the USAEC Directorate of Licensing. The preapplication review is continuing by addressing areas of concern identified in this report such that a major part of the work necessary to support the actual licensing of a GCFR demonstration plant has been established. The safety performance of the GCFR demonstration plant is based upon its inherent safety characteristics among which are the single phase and chemically inert coolant which is not activated and has a low reactivity worth, the negative core power and temperature reactivity coefficients and the small and negative steam reactivity worth. Recent studies of larger core designs indicate that as the reactor size increases central fuel, clad and coolant reactivity worths decrease and the Doppler coefficient becomes more negative. These inherent safety characteristics are complemented by safety design features such as enclosing the entire primary coolant system within a prestressed concrete pressure vessel (PCRV), providing two independent and diverse shutdown systems and residual heat removal (RHR) systems, limiting the worth of control rods to less than $1, employing pressure-equalized fuel rods, a core supported rigidly at its upper end and otherwise unrestrained and coolant downflow within the core to enhance debris removal should local melting occur. The structurally redundant PCRV design allows the potential depressurization leak area to be controlled and, since the PCRV is located within a containment building, coolant is present even after a depressurization accident and each RHR

Fluid structural response of axially cracked cylinders

International Nuclear Information System (INIS)

Garnich, M.R.; Simonen, F.A.

1985-03-01

The fluid structural (FS) response of a cylindrical pressure vessel to a suddenly occurring longitudinal through-wall crack is predicted. The effects of vessel internals and depressurization of the compressed water on dynamic crack opening displacements are investigated. A three dimensional (3D) structural finite element model is used as a basis for the development of a two dimensional (2D) FS model. A slice of the vessel taken at the crack midspan and normal to the cylinder axis is modeled. Crack opening displacements are compared between the 2D and 3D models, between the different assumptions about fluid depressurization, and between the static and dynamic solutions. The results show that effects of dynamic amplification associated with the sudden opening of the crack in the cylinder are largely offset by the local depressurization of the fluid adjacent to the crack

On the mineralization model of 'three sources--heat, water and uranium'

International Nuclear Information System (INIS)

Li Xueli

1992-01-01

In response to the relations between geological and geothermal settings, geothermal water and uranium mineralizations in the Southeastern China, the model of uranium mineralization in discharge area (depressurization area) of fossil geothermal systems in Mesozoic-Cenozoic Volcanic-magmatic active areas has been put forward and expounded in the view of mineral-formation by the 'three sources'-heat, water and uranium

Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

2016-04-15

Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

Performance behavior of the passive containment cooling system of a natural circulation BWR during postulated accident condition

International Nuclear Information System (INIS)

Kumar, Mukesh; Nayak, A.K.; Jain, Vikas; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2011-01-01

Passive systems are playing prominent role in the development of innovative nuclear reactor systems due to their simplicity, enhanced safety, reliability and economy. These systems are being considered for normal operation as well as accidental conditions of reactor following a postulated accident scenario to preclude the scenarios arising out of failure of active systems as well as to minimize the operator intervention. Indian innovative reactor AHWR being designed for thorium utilization employs various passive safety concepts. As containment is the ultimate barrier to the release of radioactivity, passive concepts are being employed in BWRs for minimize peak containment pressure in the containment during a postulated accident condition like LOCA. The concept of passive containment cooling system (PCCS) in the AHWR comprises of inclined tube heat exchangers located underneath an elevated pool that removes the heat from the steam-air atmosphere of containment following a LOCA by natural circulation of water inside the tubes. The steam condenses on the external surface of tubes of PCCS in addition to the wall of the containment which in turn depressurizes the containment. This paper deals with the performance assessment of PCCS of AHWR during a postulated design basis LOCA by using the best estimate code RELAP5/Mod3.2. (author)

ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

International Nuclear Information System (INIS)

Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

1977-01-01

ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

Advancing Ruggedness of Nuclear Stations By Expanding Defence In Depth in Critical Areas

International Nuclear Information System (INIS)

Koshy, Thomas

2015-01-01

The nuclear industry continues to rise above the challenges it has faced over the years from external events and internal events. Fukushima event has shed light on a few vulnerabilities that could be overcome by utilizing the current state of technology. Common cause from sea water ingression was not conceived to have the entire electrical power system including AC and DC disabled beyond reasonable recovery. Rather than focusing on the solutions for lessons from Fukushima, it is better to address 'Fukushima type' events and advance the resilience of the NPPs. The effort needs to be on exploring different approaches to overcome such vulnerabilities so that a variety of solutions are available to make appropriate choices on improving NPP ruggedness based on anticipated challenges in the regions. In a technology neutral approach for light water reactors (LWR) there are 4 critical areas that are significant for ensuring nuclear safety. (1) Reactor trip, (2) Depressurization, (3) Emergency Core Cooling, and (4) Containment integrity. The reactor trip had not suffered any significant setbacks in the immediate past but provisions to address Anticipated Transients without Scram (ATWS) were generally included in most designs. While the technology has advanced, software driven/assisted trips are becoming popular and desirable. However, a diverse approach with least probability of potential interference needs to be provided in the control room and remote shutdown area to advance the ruggedness of rector trip. Depressurization is essential for passive as well as active cooling systems and therefore the approaches to de-pressurize should have more than one approach to ensure its success. In the absence of diverse approaches to de-pressurize, it is more important to consider RCS cooling capability during accidents or transients while the reactor is at a higher pressure. In the area of Emergency Core Cooling, the events history demonstrates greater success on diversity

Analysis of an integrated energy system under variable loads through the symbolic exergoeconomics. 2

International Nuclear Information System (INIS)

Lazzaretto, A.; Macor, A.; Mirandola, A.; Reini, M.

1992-01-01

In this paper, the natural gas depressurization plant of the previous paper, containing a turboexpander and two cogeneration engines, is still considered. The aim of the work is to seek to what extent the Symbolic Exergoeconomic procedure can operate independently of a thermodynamic simulator. For this purpose, the results obtained by the two methodologies have to be compared when varying significant thermodynamic parameters of the system. In particular, the isentropic efficiency of the first stage turbine has been modified in a wide range. The results are very close in the hypothesis of linear variations of the symbolic parameters as a function of the turbine efficiency, whereas are not acceptable under the hypotheses of independence of the efficiencies and stability of the exergy bifurcation rations. A close correspondence between the results of the thermodynamic simulator and the symbolic procedure allows the Symbolic Exergoeconomics algorithms to be used as a basis of an optimization process. As already pointed out in previous analyses, the comparison has t be carried out for each step of the annual discretized duration curve of the gas flow rate, since the behavior of the system is very different from one step to another

Construction of the blowdown and condensation loop

Energy Technology Data Exchange (ETDEWEB)

Park, Choon Kyung; Song, Chul Kyung; Cho, Seok; Chun, S. Y.; Chung, Moon Ki

1997-12-01

The blowdown and condensation loop (B and C loop) has been constructed to get experimental data for designing the safety depressurization system (SDS) and steam sparger which are considered to implement in the Korea Next Generation Reactor (KNGR). In this report, system description on the B and C loop is given in detail, which includes the drawings and technical specification of each component, instrumentation and control system, and the operational procedures and the results of the performance testing. (author). 7 refs., 11 tabs., 48 figs.

Cooldown to residual heat removal entry conditions using atmospheric dump valves and auxiliary pressurizer spray following a loss-of-offsite power at Calvert Cliffs, Unit 1

International Nuclear Information System (INIS)

Jenks, R.P.

1984-01-01

An investigation of cooldown using atmospheric dump valves (ADVs) and auxiliary pressurizer spray (APS) following loss-of-offsite power at Calvert Cliffs-1 showed residual heat removal entry conditions could not be reached with the plant ADVs alone. Use of APS with the plant ADVs enhanced depressurization, but still provided insufficient cooldown. Effective cooldown and depressurization was shown to occur when rated steady state flow through the ADVs was increased by a factor of four. 6 refs., 30 figs., 2 tabs

Experiment data report for semiscale Mod-1 test S-04-1 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is 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 system. Test S-04-1 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using system volume scaled coolant injection parameters. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

International Nuclear Information System (INIS)

Bradley, J.G.

1982-01-01

A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception

A hybrid method of prediction of the void fraction during depressurization of diabatic systems

International Nuclear Information System (INIS)

Inayatullah, G.; Nicoll, W.B.; Hancox, W.T.

1977-01-01

The variation in vapour volumetric fraction during transient pressure, flow and power is of considerable importance in water-cooled nuclear power-reactor safety analysis. The commonly adopted procedure to predict the transient void is to solve the conservation equations using finite differences. This present method is intermediate between numerical and analytic, hence 'hybrid'. Space and time are divided into discrete intervals. Their size, however, is dictated by the imposed heat flux and pressure variations, and not by truncation error, stability or convergence, because within an interval, the solutions applied are analytic. The relatively simple hybrid method presented here can predict the void distribution in a variety of transient, diabatic, two-phase flows with simplicity, accuracy and speed. (Auth.)

THERMAL HYDRAULIC ISSUES OF CONTAINMENT FILTERED VENTING SYSTEM FOR A LONG OPERATING TIME

Directory of Open Access Journals (Sweden)

YOUNG SU NA

2014-12-01

Full Text Available This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break)

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1976-07-01

Recorded test data are presented for Test S-29-1 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 0 F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Risk evaluation of accident management strategies

International Nuclear Information System (INIS)

Dingman, S.; Camp, A.

1992-01-01

The use of Probabilistic Risk Assessment (PRA) methods to evaluate accident management strategies in nuclear power plants discussed in this paper. The PRA framework allows an integrated evaluation to be performed to give the full implications of a particular strategy. The methodology is demonstrated for a particular accident management strategy, intentional depressurization of the reactor coolant system to avoid containment pressurization during the ejection of molten debris at vessel breach

INEL design studies in support of the Westinghouse EPRI small plant study

International Nuclear Information System (INIS)

Burtt, J.D.; Kullberg, C.M.

1986-03-01

In support of the design effort of a Westinghouse EPRI small plant study, several analyses were performed at the Idaho National Engineering Laboratory. An analysis was performed to study fuel behavior under conditions of a limiting flow coastdown transient. Depressurization capabilities for the reactor coolant system were studied. The post-accident heat removal for the current containment design was studied. The results of all three studies are reported. 31 figs

Final safety evaluation report related to the certification of the System 80+ design (Docket No. 52-002). Volume 1, Chapters 1--14

International Nuclear Information System (INIS)

1994-08-01

This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE's System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report

Effect of Mineral Dissolution/Precipitation and CO2 Exsolution on CO2 transport in Geological Carbon Storage.

Science.gov (United States)

Xu, Ruina; Li, Rong; Ma, Jin; He, Di; Jiang, Peixue

2017-09-19

Geological carbon sequestration (GCS) in deep saline aquifers is an effective means for storing carbon dioxide to address global climate change. As the time after injection increases, the safety of storage increases as the CO 2 transforms from a separate phase to CO 2 (aq) and HCO 3 - by dissolution and then to carbonates by mineral dissolution. However, subsequent depressurization could lead to dissolved CO 2 (aq) escaping from the formation water and creating a new separate phase which may reduce the GCS system safety. The mineral dissolution and the CO 2 exsolution and mineral precipitation during depressurization change the morphology, porosity, and permeability of the porous rock medium, which then affects the two-phase flow of the CO 2 and formation water. A better understanding of these effects on the CO 2 -water two-phase flow will improve predictions of the long-term CO 2 storage reliability, especially the impact of depressurization on the long-term stability. In this Account, we summarize our recent work on the effect of CO 2 exsolution and mineral dissolution/precipitation on CO 2 transport in GCS reservoirs. We place emphasis on understanding the behavior and transformation of the carbon components in the reservoir, including CO 2 (sc/g), CO 2 (aq), HCO 3 - , and carbonate minerals (calcite and dolomite), highlight their transport and mobility by coupled geochemical and two-phase flow processes, and consider the implications of these transport mechanisms on estimates of the long-term safety of GCS. We describe experimental and numerical pore- and core-scale methods used in our lab in conjunction with industrial and international partners to investigate these effects. Experimental results show how mineral dissolution affects permeability, capillary pressure, and relative permeability, which are important phenomena affecting the input parameters for reservoir flow modeling. The porosity and the absolute permeability increase when CO 2 dissolved water is

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)

Validation of thermohydraulic codes by comparison of experimental results with computer simulations

International Nuclear Information System (INIS)

Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

1989-01-01

The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 1, Chapters 1--14

Energy Technology Data Exchange (ETDEWEB)

1994-08-01

This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE`s System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report.

The possible role of indoor radon reduction systems in back-drafting residential combustion appliances

International Nuclear Information System (INIS)

Henschel, D.B.

1997-01-01

A computational sensitivity analysis was conducted to identify the conditions under which residential active soil depressurization (ASD) systems for indoor radon reduction might most likely exacerbate or create back-drafting of natural-draft combustion appliances. Parameters varied included: house size; normalized leakage area; exhaust rate of exhaust appliances other than the ASD system; and the amount of house air exhausted by the ASD system. Even with a reasonably conservative set of assumptions, it is predicted that ASD systems should not exacerbate or create back- drafting in most of the U.S. housing stock. Only at normalized leakage areas lower than 3 to 4 cm 2 commercial at 4 Pa) per m 2 of floor area should ASD contribute to back-drafting, even in small houses at high ASD exhaust rates (compared to a mean of over 10 cm 2 /m 2 determined from data on over 12,000 U.S. houses). But on the other hand, even with a more forgiving set of assumptions, it is predicted that ASD systems could contribute to back-drafting in some fraction of the housing stock -houses tighter than about 1 to 2 cm 2 /m 2 - even in large houses at minimal ASD exhaust rates. It is not possible to use parameters such as house size or ASD system flow rate to estimate reliably the risk that an ASD system might contribute to back-drafting in a given house. Spillage/back-draft testing would be needed for essentially all installations. (au) 18 refs

Experimental study of aerosol reentrainment from flashing pool in ALPHA program

International Nuclear Information System (INIS)

Kudo, T.; Yamano, N.; Moriyama, K.; Maruyama, Y.; Sugimoto, J.

1994-01-01

Aerosol reentrainment experiments are being performed as a part of the ALPHA (Assessment of Loads and Performance of Containment in a Hypothetical Accident) program at JAERI (Japan Atomic Energy Research Institute). The major objective of the experiments is to quantify and characterize the reentrainment of the dissolved material from a flashing pool during the rapid depressurization of a reactor containment vessel. Two experiments were performed. In the experiments a water pool dissolving sodium sulfate as FP simulant was located in the model containment vessel and the containment breach area was simulated with an orifice with 24 mm diameter. This orifice was estimated to give the same order of depressurization rate as the case of BWR Mark 1 containment failure with most likely breach size. In the first experiment ARE001, a pool water of 800 kg dissolving 50 kg of sodium sulfate was employed. The model containment was depressurized from 1.5 MPa to 0.1 MPa in approximately 45 minutes. In the second experiment ARE002, the mass of the pool water was reduced to 400 kg dissolving 25 kg of sodium sulfate. The internal pressure of the containment was decreased from 1.3 MPa to 0.1 MPa in approximately 40 minutes. At the beginning of the depressurization the pool water was heated to the saturation temperature at the internal pressure of the containment. The entrained droplets were sampled during depressurization period. Sodium sulfate deposited in all parts of the test facility was collected and weighed after the experiments. Results of the experiments showed that very small fraction of the dissolved material (less than 0.03%) was reentrained although approximately, 20% of water was evaporated from the pool water. The reentrained mass predicted with the Kataoka-Ishii model was approximately 1/110 of the mass evaluated in the experiments. This may be due to multi-dimensional features of the pool geometry. (author)

Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

Energy Technology Data Exchange (ETDEWEB)

Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

Directory of Open Access Journals (Sweden)

HEOK-SOON LIM

2014-02-01

Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

Energy Technology Data Exchange (ETDEWEB)

Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

2014-02-15

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

International Nuclear Information System (INIS)

Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo; Kim, Seoungrae

2014-01-01

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident

Design, construction and characterization of pneumatic system for measurement of roughness

Directory of Open Access Journals (Sweden)

Germán A. Bacca-Bastidas

2015-07-01

Full Text Available This article aims to present the results obtained from the design, construction and characterization of a pneumatic flapper-nozzle amplifier, employed in the measurement of average surface roughness. In the construction of the sensor, low cost materials were used and most pieces were obtained by machining. The data acquisition was performed through PC, using an Arduino interface board. The nonlinear mathematical model of the sensor is based on equations of perfect gas flow through an orifice and the continuity law for a control volume. The characterization of physical parameters obtained through laboratory techniques based on the transient response of the gas pressure in the pressurization and depressurization processes of constant volume chambers, using computational tools for adjusting experimental curves. The validation of the model was based on the specifications of transient response that presents a dynamic system for a step input. For the measurement of the roughness, the mathematical model of average roughness, Ra, was used, and the measured data by the sensor were obtained in sandpapers from P1000 to P2000 size, with reference for validation values of average roughness indicated by the FEPA standard.

Final safety evaluation report related to the certification of the System 80+ design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices

International Nuclear Information System (INIS)

1994-08-01

This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE's System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E

CFD Analyses of Air-Ingress Accident for VHTRs

Science.gov (United States)

Ham, Tae Kyu

-ingress mechanism and to utilize the CFD simulation in the analysis of the phenomenon. Previous air-ingress studies simulated the depressurization process using simple assumptions or 1-D system code results. However, recent studies found flow oscillations near the end of the depressurization which could influence the next stage of the air-ingress accident. Therefore, CFD simulations were performed to examine the air-ingress mechanisms from the depressurization through the establishment of local natural circulation initiate. In addition to the double-guillotine break scenario, there are other scenarios that can lead to an air-ingress event such as a partial break were in the cross vessel with various break locations, orientations, and shapes. These additional situations were also investigated. The simulation results for the OSU test facility showed that the discharged helium coolant from a reactor vessel during the depressurization process will be mixed with the air in the containment. This process makes the density of the gas mixture in the containment lower and the density-driven air-ingress flow slower because the density-driven flow is established by the density difference of the gas species between the reactor vessel and the containment. In addition, for the simulations with various initial and boundary conditions, the simulation results showed that the total accumulated air in the containment collapsed within 10% standard deviation by: 1. multiplying the density ratio and viscosity ratio of the gas species between the containment and the reactor vessel and 2. multiplying the ratio of the air mole fraction and gas temperature to the reference value. By replacing the gas mixture in the reactor cavity with a gas heavier than the air, the air-ingress speed slowed down. Based on the understanding of the air-ingress phenomena for the GT-MHR air-ingress scenario, several mitigation measures of air-ingress accident are proposed. The CFD results are utilized to plan experimental

Mechanical model of human eye compliance for volumetric occlusion break surge measurements.

Science.gov (United States)

Dyk, David W; Miller, Kevin M

2018-02-01

To develop a mechanical model of human eye compliance for volumetric studies. Alcon Research, Ltd., Lake Forest, California, USA. Experimental study. Enucleated human eyes underwent pressurization and depressurization cycles with peak intraocular pressures (IOPs) of 60 to 100 mm Hg; anterior chamber pressure and volume changes were measured. Average net volume change curves were calculated as a function of IOP for each eye. Overall mean volumes were computed from each eye's average results at pressure points extrapolated over the range of 5 to 90 mm Hg. A 2-term exponential function was fit to these results. A fluid chamber with a displaceable piston was created as a mechanical model of this equation. A laser confocal displacement meter was used to measure piston displacement. A test bed incorporated the mechanical model with a mounted phacoemulsification probe and allowed for simulated occlusion breaks. Surge volume was calculated from piston displacement. An exponential function, V = C 1 × exp(C 2 × IOP) + C 3  × exp(C 4  × IOP) - V 0 , where V, the volume, was fit to the final depressurization curve obtained from 15 enucleated human eyes. The C 1 through C 4 values were -0.07141, -0.23055, -0.14972, and -0.02006, respectively. The equation was modeled using a piston system with 3 parallel springs that engaged serially. The mechanical model mimicked depressurization curves observed in human cadaver eyes. The resulting mechanical compliance model measured ocular volumetric changes and thus would be helpful in characterizing the postocclusion break surge response. Copyright © 2018 ASCRS and ESCRS. Published by Elsevier Inc. All rights reserved.

Dependence of radon level on ventilation systems in residences

International Nuclear Information System (INIS)

Kokotti, H.

1995-01-01

The concentration of indoor radon and radon entry from soil into a house are expected to increase with increasing radon concentration in soil pores, and indoor radon concentration is expected to decrease with increasing ventilation rate. Depressurization, which can be caused by the stack effect, by wind and by unbalanced ventilation, creates different pressure conditions in a house and in the soil beneath it. To reveal the possible differences in radon removal and entry resulting from different ventilation systems, radon concentrations were determined in three similar slab-on-grade buildings provided with mechanical supply and exhaust ventilation, mechanical exhaust or natural ventilation. To limitate the effect of differences in soil parameters, the houses were constructed on the same gravel esker in Kuopio. Thus, the variation in radon entry as a result of different depressurisation of the houses (caused by unbalanced mechanical ventilation systems) could also be observed. In addition, the effect of pressurisation on living rooms could be determined in five slab-on-grade houses constructed on the same esker in Hollola. Mechanical supply and exhaust ventilation system controlled by measured indoor-outdoor pressure difference, was installed in the six houses. The seasonal variation with and without controlled pressure conditions were followed in a slab-on-grade house constructed on a gravel esker in Rekola. Long-term radon concentrations were observed to correlate negatively with air exchange rates. However, the removal effect of ventilation was found to be disturbed by negative pressure due to the stack effect and/or to unbalanced mechanical ventilation. (91 refs., 17 figs., 10 tabs.)

Pitot tube and drag body measurements in transient steam--water flows

International Nuclear Information System (INIS)

Fincke, J.R.; Deason, V.A.; Dacus, M.W.

1979-01-01

The use of full-flow drag devices and rakes of water-cooled Pitot tubes to measure the transient two-phase mass flow during loss-of-coolant experiments in pressurized water reactor (PWR) environments has been developed. Mass flow rate measurements have been obtained in high temperature and pressure environments, similar to PWRs, under transient conditions. Comparisons of the measured time integrated value of mass flow to the known system mass before depressurization are made

Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices

Energy Technology Data Exchange (ETDEWEB)

1994-08-01

This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE`s System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E.

Experimental investigations on the deposition and remobilization of aerosol particles in turbulent flows

International Nuclear Information System (INIS)

Barth, Thomas

2014-01-01

Aerosol particle deposition and resuspension experiments in turbulent flows were performed to investigate the complex particle transport phenomena and to provide a database for the development and validation of computational fluid dynamics (CFD) codes. The background motivation is related to the source term analysis of an accidental depressurization scenario of a High Temperature Reactor (HTR). During the operation of former HTR pilot plants, larger amounts of radio-contaminated graphite dust were found in the primary circuit. This dust most likely arose due to abrasion between the graphitic core components and was deposited on the inner wall surfaces of the primary circuit. In case of an accident scenario, such as a depressurization of the primary circuit, the dust may be remobilized and may escape the system boundaries. The estimation of the source term being discharged during such a scenario requires fundamental knowledge of the particle deposition, the amount of contaminants per unit mass as well as the resuspension phenomena. Nowadays, the graphite dust distribution in the primary circuit of an HTR can be calculated for stationary conditions using one-dimensional reactor system codes. However, it is rather unknown which fraction of the graphite dust inventory may be remobilized during a depressurization of the HTR primary circuit. Two small-scale experimental facilities were designed and a set of experiments was performed to investigate particle transport, deposition and resuspension in turbulent flows. The facility design concept is based on the fluid dynamic downscaling of the helium pressure boundary in the HTR primary circuit to an airflow at ambient conditions in the laboratory. The turbulent flow and the particles were recorded by high-resolution, non-invasive imaging techniques to provide a spatio-temporal insight into the particle transport processes. The different investigations of this thesis can be grouped into three categories. Firstly, the

Development and industrial tests of the first LNG hydraulic turbine system in China

Directory of Open Access Journals (Sweden)

Jie Chen

2016-10-01

Full Text Available The cryogenic hydraulic turbine can be used to replace the conventional J–T valve for LNG or mixed refrigerant throttling and depressurization in a natural gas liquefaction plant. This advanced technology is not only to enhance the efficiency of the liquefaction plant, but to usher a new trend in the development of global liquefaction technologies. China has over 136 liquefaction plants, but the cryogenic hydraulic turbines have not been deployed in industrial utilization. In addition, these turbines cannot be manufactured domestically. In this circumstance, through working on the key technologies for LNG hydraulic turbine process & control system development, hydraulic model optimization design, structure design and manufacturing, the first domestic cryogenic hydraulic turbine with a flow rate of 40 m3/h was developed to recover the pressure energy from the LNG of cold box. The turbine was installed in the CNOOC Zhuhai Natural Gas Liquefaction Plant for industrial tests under multiple working conditions, including start-stop, variable flow rates and variable rotation speeds. Test results show that the domestic LNG cryogenic hydraulic turbine has satisfactory mechanical and operational performances at low temperatures as specified in design. In addition, the process & control system and frequency-conversion power-generation system of the turbine system are designed properly to automatically and smoothly replace the existing LNG J–T valve. As a result, the domestic LNG cryogenic hydraulic turbine system can improve LNG production by an average of 2% and generate power of 8.3 kW.

Dynamic pore network simulator for modelling buoyancy-driven migration during depressurisation of heavy-oil systems

Energy Technology Data Exchange (ETDEWEB)

Ezeuko, C.C.; McDougall, S.R. [Heriot-Watt Univ., Edinburgh (United Kingdom); Bondino, I. [Total E and P UK Ltd., London (United Kingdom); Hamon, G. [Total S.A., Paris (France)

2008-10-15

In an attempt to investigate the impact of gravitational forces on gas evolution during solution gas drive, a number of vertically-oriented heavy oil depletion experiments have been conducted. Some of the results of these studies suggest the occurrence of gas migration during these tests. However, a major limitation of these experiments is the difficulty in visualizing the process in reservoir rock samples. Experimental observations using transparent glass models have been useful in this context and provide a sound physical basis for modelling gravitational gas migration in gas-oil systems. This paper presented a new pore network simulator that was capable of modelling the time-dependent migration of growing gas structures. Multiple pore filling events were dynamically modelled with interface tracking allowing the full range of migratory behaviours to be reproduced, including braided migration and discontinuous dispersed flow. Simulation results were compared with experiments and were found to be in excellent agreement. The paper presented the model and discussed the implication of evolution regime on recovery from heavy oil systems undergoing depressurization. The simulation results demonstrated the complex interaction of a number of network and fluid parameters. It was concluded that the concomitant effect on the competition between capillarity and buoyancy produced different gas evolution patterns during pressure depletion. 28 refs., 2 tabs., 19 figs.

Development of risk-informed system design methodology for future nuclear power plants

International Nuclear Information System (INIS)

Yang, J. H.; Kim, M. R.; Park, S. J.; Lim, H. K.; Ji, S. K.; Choi, C. J.

2002-01-01

The purpose of this analysis is to develop the risk assessment evaluation process that can reduce the conservatism involved in the LOCA quantification. The frequency estimation for LOCA was performed according to NUREG/CR-5750. The raw data for LOCA events described in NUREG/CR-5750 was applied to this project. Lots of thermal hydraulic analyses for various break sizes were performed to find the boundary conditions that can effect the success criteria of event mitigation. The MARS 2.1 code, best-estimated computer code, was used in this analysis. The analysis result shows that conservatism in the LOCA quantification can be reduced when the detailed LOCA breakdown supported thermal hydraulic analysis is performed in the PSA model. The CDF for new re-classified LOCA events was reduced about 50% of current model's. Concurrent with the LOCA re-classification, the operator's available time for the feed and bleed operation using Safety Depressurization System (SDS) valves during small LOCA and its contribution to CDF were considered. Its results did not have an effect of CDF reduction, but it is believed that the iterative approach and findings are very useful

Control system for the feed of pressurized fluid in a hydraulic circuit as a function of the state of the locking or unlocking of two mechanical organs

International Nuclear Information System (INIS)

Huet, Y.; Perichon, C.

1985-01-01

The control system comprises two hydraulic cylinders of which rods are integral with the mechanical organs. The piston of the first cylinder separates the chamber of this one in two parts. The piston of the second cylinder separates its chamber in three parts. The inlet chamber of the two cylinders are connected to pressurized fluid feed pipes, and the outlet chambers to a depressurization pipe. According to the position of the piston depending itself on the state of locking or unlocking of the rods, an interconnection pipe and a feed pipe of the pressurized fluid hydraulic circuit communicate with a chamber or another one. The feed of the hydraulic circuit is possible only the two rods are unlocked. The invention applies more particularly to the feed of the control circuit of an emergency seal of the primary pump of a pressurized water nuclear reactor [fr

Physics related to control and safety of hybrid systems; Physique associee au controle et a la surete des systemes hybrides

Energy Technology Data Exchange (ETDEWEB)

Gueton, O

2001-12-01

Regarding nuclear waste management, ADS can be considered as large minor actinides burners. In a first part, a critical analysis of different reactor types shows that fast spectrum, helium coolant and nitride fuel, containing 100% minor actinides, agree perfectly with the high transmutation requirements of ADS. The control and safety demonstration of this system represents the main purpose of this study. Understanding spatial and dynamic behaviour of ADS flux is absolutely necessary. For this purpose, we have defined an indicator to quantify spatial decoupling. It shows, on the one hand, point kinetic deficiency to study local transients, and on the other hand, perturbations propagation differences between ADS and critical cores. Then, in a more concrete approach, accidental sequences (source transient, beam de-focalization, reactivity insertions, loss of flow, depressurization) are evaluated for this core, strongly loaded with minor actinides. It is shown that the automatic beam shutdown leads to preserve large safety margins for all studied transients. The accelerator emergency stop is induced by an unexpected evolution of the core control parameters. These parameters, except reactivity, can be directly measured in subcritical systems like in critical ones. Concerning reactivity, we suggest a new method for its absolute determination in ADS: at the time of reactor start-up, the reactivity must be calibrated by coupling two methods of relative reactivity measurements (pulsed source and Approached Source Multiplication) for successive subcritical levels. After that, the on-line follow-up of reactivity is obtained from this calibration like in a critical core. (authors)

The Interplay Between Saline Fluid Flow and Dynamic Permeability in Magmatic-Hydrothermal Systems

Science.gov (United States)

Weis, P.

2014-12-01

Magmatic-hydrothermal ore deposits document the interplay between saline fluid flow and rock permeability. Numerical simulations of multi-phase flow of variably miscible, compressible H20-NaCl fluids in concert with a dynamic permeability model can reproduce characteristics of porphyry copper and epithermal gold systems. This dynamic permeability model incorporates depth-dependent permeability profiles characteristic for tectonically active crust as well as pressure- and temperature-dependent relationships describing hydraulic fracturing and the transition from brittle to ductile rock behavior. In response to focused expulsion of magmatic fluids from a crystallizing upper crustal magma chamber, the hydrothermal system self-organizes into a hydrological divide, separating an inner part dominated by ascending magmatic fluids under near-lithostatic pressures from a surrounding outer part dominated by convection of colder meteoric fluids under near-hydrostatic pressures. This hydrological divide also provides a mechanism to transport magmatic salt through the crust, and prevents the hydrothermal system to become "clogged" by precipitation of solid halite due to depressurization of saline, high-temperature magmatic fluids. The same physical processes at similar permeability ranges, crustal depths and flow rates are relevant for a number of active systems, including geothermal resources and excess degassing at volcanos. The simulations further suggest that the described mechanism can separate the base of free convection in high-enthalpy geothermal systems from the magma chamber as a driving heat source by several kilometers in the vertical direction in tectonic settings with hydrous magmatism. This hydrology would be in contrast to settings with anhydrous magmatism, where the base of the geothermal systems may be closer to the magma chamber.

Cascade Storage and Delivery System for a Multi Mission Space Exploration Vehicle (MMSEV)

Science.gov (United States)

Yagoda, Evan; Swickrath, Michael; Stambaugh, Imelda

2012-01-01

NASA is developing a Multi Mission Space Exploration Vehicle (MMSEV) for missions beyond Low Earth Orbit (LEO). The MMSEV is a pressurized vehicle used to extend the human exploration envelope for Lunar, Near Earth Object (NEO), and Deep Space missions. The Johnson Space Center is developing the Environmental Control and Life Support System (ECLSS) for the MMSEV. The MMSEV s intended use is to support longer sortie lengths with multiple Extra Vehicular Activities (EVAs) on a higher magnitude than any previous vehicle. This paper presents an analysis of a high pressure oxygen cascade storage and delivery system that will accommodate the crew during long duration Intra Vehicular Activity (IVA) and capable of multiple high pressure oxygen fills to the Portable Life Support System (PLSS) worn by the crew during EVAs. A cascade is a high pressure gas cylinder system used for the refilling of smaller compressed gas cylinders. Each of the large cylinders are filled by a compressor, but the cascade system allows small cylinders to be filled without the need of a compressor. In addition, the cascade system is useful as a "reservoir" to accommodate low pressure needs. A regression model was developed to provide the mechanism to size the cascade systems subject to constraints such as number of crew, extravehicular activity duration and frequency, and ullage gas requirements under contingency scenarios. The sizing routine employed a numerical integration scheme to determine gas compressibility changes during depressurization and compressibility effects were captured using the Soave-Redlich-Kwong (SRK) equation of state. A multi-dimensional nonlinear optimization routine was used to find the minimum cascade tank system mass that meets the mission requirements. The sizing algorithms developed in this analysis provide a powerful framework to assess cascade filling, compressor, and hybrid systems to design long duration vehicle ECLSS architecture. 1

Recent progress of China HCCB TBM tritium system

Energy Technology Data Exchange (ETDEWEB)

Luo, Deli, E-mail: luodeli2005@hotmail.com; Huang, Guoqiang; Huang, Zhiyong; Qin, Cheng; Song, Jiangfeng; He, Kanghao; Chen, Chang’an; Zhang, Guikai; Fu, Jun; Yao, Yong; An, Yongtao

2016-11-01

Highlights: • Comparing with our previous design, improvements have been made according to the up-to-date experiments and simulations: (1) The palladium alloy tube in the previous design is now removed in the upgraded one and the cryogenic molecular sieve bed is replaced by the getter bed to reduce tritium inventory; (2) Hot metal reduction bed is relocated from T-Plant to Port Cell; (3) TAS is now integrated into TES. • The proposed coolant purification is based on catalytic oxidation and molecular sieve bed adsorption for tritium removal, as well as hot metal adsorption for the elimination of non-tritium gaseous impurities. Some operation parameters and functional components are improved. The interface with the high pressure HCS and other plant systems was incorporated taking into account of the requirement from the ITER port management group meetings. - Abstract: China tritium system including Tritium Extraction System (TES) with Tritium Accountancy System (TAS) integrated in and Coolant Purification System (CPS), which is subordinate to Helium Coolant System (HCS), is of great importance for China Helium Cooled Ceramic Breeder Test Blanket Module (CN HCCB TBM). The purge gas (99.9% He + 0.1% H{sub 2}) carrying Q{sub 2}O (Q = H, D, T) and Q{sub 2} from Li{sub 4}SiO{sub 4} ceramic breeder flows through the reduction bed where Q{sub 2}O is reduced into Q{sub 2} and then absorbed by the getter bed. The HT/HTO ratio and the total tritium are determined by TAS. Catalytic oxidation combines with molecular sieve absorption and hot metal purification are applied to remove tritium and other impurities in helium coolant. A loop including depressurization, helium-sweeping assisted thermal desorption, and cold trapping for the regeneration of saturated molecular sieve bed until the concentration of the desorbed Q{sub 2}O is reduced to an acceptable level. This paper introduces the recent progress of China tritium system including updated conceptual designs of TES and

Critical pressure of non-equilibrium two-phase critical flow

Energy Technology Data Exchange (ETDEWEB)

Minzer, U [Israel Electric Corp. Ltd., Haifa (Israel)

1996-12-01

Critical pressure is defined as the pressure existing at the exit edge of the piping, when it remains constant despite a decrease in the back. According to this definition the critical pressure is larger than the back pressure and for two-phase conditions below saturation pressure. The two-phase critical pressure has a major influence on the two-phase critical flow characteristics. Therefore it is of High significance in calculations of critical mass flux and critical depressurization rate, which are important in the fields of Nuclear Reactor Safety and Industrial Safety. At the Nuclear Reactor Safety field is useful for estimations of the Reactor Cooling System depressurization, the core coolant level, and the pressure build-up in the containment. In the Industrial Safety field it is helpful for estimating the leakage rate of toxic gases Tom liquefied gas pressure vessels, depressurization of pressure vessels, and explosion conditions due to liquefied gas release. For physical description of non-equilibrium two-phase critical flow it would be convenient to divide the flow into two stages. The first stage is the flow of subcooled liquid at constant temperature and uniform pressure drop (i.e., the case of incompressible fluid and uniform piping cross section). The rapid flow of the liquid causes a delay in the boiling of the liquid, which begins to boil below saturation pressure, at thermal non-equilibrium. The boiling is the beginning of the second stage, characterized by a sharp increase of the pressure drop. The liquid temperature on the second stage is almost constant because most of the energy for vaporization is supplied from the large pressure drop The present work will focus on the two-phase critical pressure of water, since water serves as coolant in the vast majority of nuclear power reactors throughout the world. (author).

Critical pressure of non-equilibrium two-phase critical flow

International Nuclear Information System (INIS)

Minzer, U.

1996-01-01

Critical pressure is defined as the pressure existing at the exit edge of the piping, when it remains constant despite a decrease in the back. According to this definition the critical pressure is larger than the back pressure and for two-phase conditions below saturation pressure. The two-phase critical pressure has a major influence on the two-phase critical flow characteristics. Therefore it is of High significance in calculations of critical mass flux and critical depressurization rate, which are important in the fields of Nuclear Reactor Safety and Industrial Safety. At the Nuclear Reactor Safety field is useful for estimations of the Reactor Cooling System depressurization, the core coolant level, and the pressure build-up in the containment. In the Industrial Safety field it is helpful for estimating the leakage rate of toxic gases Tom liquefied gas pressure vessels, depressurization of pressure vessels, and explosion conditions due to liquefied gas release. For physical description of non-equilibrium two-phase critical flow it would be convenient to divide the flow into two stages. The first stage is the flow of subcooled liquid at constant temperature and uniform pressure drop (i.e., the case of incompressible fluid and uniform piping cross section). The rapid flow of the liquid causes a delay in the boiling of the liquid, which begins to boil below saturation pressure, at thermal non-equilibrium. The boiling is the beginning of the second stage, characterized by a sharp increase of the pressure drop. The liquid temperature on the second stage is almost constant because most of the energy for vaporization is supplied from the large pressure drop The present work will focus on the two-phase critical pressure of water, since water serves as coolant in the vast majority of nuclear power reactors throughout the world. (author)

Depressurized pipes decontamination by using circulation foam

International Nuclear Information System (INIS)

Damerval, Frederique; Belz, Jacques; Renouf, Marjorie; Janneau, Patrice

2012-09-01

Decontamination of pipes remains a necessity in order to reduce the radiation level during maintenance or dismantling operations but it is not so easy to do it, especially in case of a long pipe network. To achieve this operation, the use of chemistry is one of the more relevant methods; moreover, the liquid waste production still remains an issue that it can be avoided by the use of decontamination foams. (authors)

Testing of indoor radon reduction techniques in central Ohio houses: Phase 2 (Winter 1988-1989). Final report, September 1988-May 1989

International Nuclear Information System (INIS)

Findlay, W.O.; Robertson, A.; Scott, A.G.

1990-05-01

The report gives results of tests of developmental indoor radon reduction techniques in nine slab-on-grade and four crawl-space houses near Dayton, Ohio. The slab-on-grade tests indicated that, when there is a good layer of aggregate under the slab, the sub-slab ventilation (SSV) mitigation technique, with only one or two suction pipes, can generally reduce indoor concentrations below 2 pCi/L (86 to 99% reduction). These reductions can be achieved even when: there are forced-air supply ducts under the slab; the slab is large (up to 2600 sq ft); and the foundation walls are hollow block. Operating the SSV system in suction always gave greater reductions than did operating in pressure. The crawl-space tests demonstrated that depressurizing under a plastic liner over the crawl-space floor was able to reduce living-area radon concentrations below 2 pCi/L (81 to 96% reduction). The performance of such sub-liner depressurization gave better reductions than did crawl-space ventilation (blowing air into, or out of, the crawl space). Completely covering the crawl-space floor with plastic sheeting was not always necessary to get adequate performance

A study of the influence of a gravel subslab layer on radon entry rate using two basement structures

International Nuclear Information System (INIS)

Robinson, A.L.; Sextro, R.G.; Fisk, W.J.; Garbesi, K.; Wooley, J.; Wollenberg, H.A.

1993-01-01

In buildings with elevated radon concentrations, the dominant transport mechanism of radon is advective flow of soil gas into the building substructure. However, the building-soil system is often complex, making detailed studies of the radon source term difficult. In order to examine radon entry into buildings, the authors have constructed two room-size, precisely-fabricated basement structures at a site with relatively homogeneous, moderately permeable soil. The basements are identical except that one lies directly on native soil whereas the other lies on a high permeability aggregate layer. The soil pressure field and radon entry rate have been measured for different basement pressures and environmental conditions. The subslab gravel layer greatly enhances the advective entry of radon into the structure; when the structures are depressurized, the radon entry rate into the structure with the subslab gravel layer is more than a factor of 3 times the radon entry rate into the other structure for the same depressurization. The gravel subslab layer also spreads the pressure field around the structure, extending the field of influence of the structure and the region from which it draws radon

A study of the influence of a gravel subslab layer on radon entry rate using two basement structures

International Nuclear Information System (INIS)

Robinson, A.L.; Sextro, R.G.; Fisk, W.J.; Garbesi, K.; Wooley, J.; Wollenberg, H.A.

1993-01-01

In buildings with elevated radon concentrations, the dominant transport mechanism of radon is advective flow of soil gas into the building substructure. However, the building-soil system is often complex, making detailed studies of the radon source term difficult. In order-to examine radon entry into buildings, we have constructed two room-size, precisely-fabricated basement structures at a site with relatively homogeneous, moderately permeable soil. The basements are identical except that one lies directly on native soil whereas the other lies on a high permeability aggregate layer. The soil pressure field and radon entry rate have been measured for different basement pressures and environmental conditions. The subslab gravel layer greatly enhances the advective entry of radon into the structure; when the structures are depressurized, the radon entry rate into the structure with the subslab gravel layer is more than a factor of 3 times the radon entry rate into the other structure for the same depressurization. The gravel subslab layer also spreads the pressure field around the structure, extending the field of influence of the structure and the region from which it draws radon. (orig.). (7 refs., 3 figs.)

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.)

Analysis on engineering application of CNP1000 in-containment refueling water storage tank

International Nuclear Information System (INIS)

Wang Bin; Wang Yong; Qiu Jian; Weng Minghui

2005-01-01

Based on the basic design of CNP1000 (three loops), which is self-reliance designed by China National Nuclear Cooperation, and investigation results from abroad advanced nuclear power plant design of In-containment Refueling Water Storage tank, this paper describe the system flowsheet, functional requirements, structural design and piping arrangement about In-containment Refueling Water Storage Tank. The design takes the lower structural space as the IRWST. Four areas are configured to meet the diverse functional requirements, including depressurization area, water collection area, safety injection and/or containment spray suction area, TSP storage area / reactor cavity flooding holdup tank. Also the paper depict the corresponding analysis and demonstration, such as In-containment Refueling Water Storage Tank pressure transient on depressurization area of IRWST, suction and internal flow stream of IRWST, configuration of strains, the addition method and amount of chemical addition, design and engineering applicant of Reactor Cavity Flooding System. All the analysis results show the basic design of IRWST meeting with the Utility Requirement Document's requirements on performance of safety function, setting of overfill passage, overpressure protection, related interference, etc., and show the reliability of Engineering Safety Features being improved for CNP1000 (three loops). Meanwhile, it is demonstrated that the design of In-containment Refueling Water Storage Tank can apply on the future nuclear power plant project in China. (authors)

Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

Energy Technology Data Exchange (ETDEWEB)

Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2017-05-15

Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

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)

Experiment data report for semiscale Mod-1 test S-01-1 (isothermal blowdown with core resistance simulator)

International Nuclear Information System (INIS)

Zender, S.N.; Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1975-04-01

Recorded test data are presented for Test S-01-1 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-1 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is representative of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-1 was conducted from an initial temperature of 540 0 F and an initial pressure of 1596 psig. A simulated intermediate size double-ended hot leg break (0.00145 ft 2 break area on each end) was used to investigate the system response to a slow depressurization transient. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. During system depressurization, coolant was injected into the vessel downcomer inlet annulus to investigate the effectiveness of injection into the inlet annulus with respect to delivery of coolant to the lower plenum. Following the blowdown portion of Test S-01-1, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The purpose of this report is to make available the uninterpreted data from Test S-01-1 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent. (U.S.)

Recommendations of the Reaktorsicherheitskommission, adopted at the 226th meeting on October 21, 1987

International Nuclear Information System (INIS)

1988-01-01

Internal emergency protection covers all measures that use in a flexible manner the existing operating and safety systems in the plant, for controlling or limiting design-exceeding incidents, and the consequences of such incidents. The following additional measures are planned or already installed: - Depressurization of the containment through a filter system. - Filtering of control room supply air. - Discharge time of the batteries of the emergency power system to be between 2 and 3 hours. - Use of the turbo-generator driven feed system of the BWR design type 69, in case of assumed, simultaneous failure of the auxiliary electrical system and of the diesel emergency sets. - Inerting of the containment of BWR of the 69 design type. (orig./HP) [de

Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP

International Nuclear Information System (INIS)

Vargas O, D.; Chavez M, C.

2012-10-01

A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

Test Methods and Protocols for Environmental and Safety Hazards Associated with Home Energy Retrofits

Energy Technology Data Exchange (ETDEWEB)

Cautley, D. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Viner, J. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Lord, M. [NorthernSTAR Building America Partnership, St. Paul, MN (United States); Pearce, M. [NorthernSTAR Building America Partnership, St. Paul, MN (United States)

2012-12-01

A number of health hazards and hazards to the durability of homes may be associated with energy retrofitting and home renovation projects. Among the hazards associated with energy retrofit work, exposure to radon is thought to cause more than 15,000 deaths per year in the U.S., while carbon monoxide poisoning results in about 20,000 injuries and 450 deaths per year. Testing procedures have been developed for identifying and quantifying hazards during retrofitting. These procedures commonly include a battery of tests to screen combustion appliances for safe operation, including worst case depressurization measurement, backdrafting (spillage) under depressurized or normal conditions, and carbon monoxide production.

Thermal hydraulic research on next generation PWRs using ROSA/LSTF

International Nuclear Information System (INIS)

Yonomoto, T.; Anoda, Y.

2000-01-01

A thermal-hydraulic research on next generation PWRs has been conducted at JAERI using the ROSA-V/Large Scale Test Facility (LSTF), focusing on phenomena related to passive safety systems. This paper describes two test results conducted for this research: a small break loss-of-coolant accident (SBLOCA) test and a low pressure steady-state natural circulation (NC) test. The former test investigated a combined use of a SG secondary-side automatic depressurization system (SADS) and a gravity-driven injection system (GDIS) to mitigate a SBLOCA. The results have shown that the primary loop can be depressurized to the GDIS actuation pressure of 0.2 MPa by the SADS alone, and then the stable long-term core cooling can be established by NC. Results of both tests showed a complicated nonuniform flow behavior among SG U-tubes during NC, which was characterized by the coexistence of concurrent condensing two-phase flow in some tubes and stagnant two-phase stratification in the others. The mechanism for the stratification was understood from the measured secondary side temperature distribution showing the lowest temperature at the top and bottom regions and the highest around the midplane. This was caused by the saturation temperature difference corresponding to the static pressure difference, and the recirculation in the secondary. This secondary side temperature distribution enabled the condensation occurring around the tube top to be balanced with evaporation occurring around the midplane in the U-tube with the stratification. Since the heat transfer occurs primarily through tubes with the concurrent flow, the nonuniform behavior directly affects the effective heat transfer area at SG. When the SG primary side was modeled with one lumped flow channel, the RELAP5 significantly over predicted the primary depressurization rate, and could not predict the stable long-term core cooling behavior at low pressure. In order to understand the mechanism of the nonuniform behavior, the

Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

International Nuclear Information System (INIS)

Hadid Subki, M.; )

2014-01-01

The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

International Nuclear Information System (INIS)

Cao Hongjun; Yan Xiuping

2013-01-01

The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-02

International Nuclear Information System (INIS)

Kukita, Yutaka; Hirata, Kazuo; Gotou, Hiroki

1990-03-01

Experimental data for the 10% hot leg break test, Run SB-HL-02, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on June 30, 1987, are presented. This test assumed total failure of both high pressure injection (HPI) and auxiliary feedwater (AFW) systems. The test results were characterized by asymmetric loop responses, flashing in the cold legs and upper downcomer, as well as condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

Response of ventilation dampers to large airflow pulses

International Nuclear Information System (INIS)

Gregory, W.S.; Smith, P.R.

1985-04-01

The results of an experiment program to evaluate the response of ventilation system dampers to simulated tornado transients are reported. Relevant data, such as damper response time, flow rate and pressure drop, and flow/pressure vs blade angle, were obtained, and the response of one tornado protective damper to simulated tornado transients was evaluated. Empirical relationships that will allow the data to be integrated into flow dynamics codes were developed. These flow dynamics codes can be used by safety analysts to predict the response of nuclear facility ventilation systems to tornado depressurization. 3 refs., 21 figs., 6 tabs

Measurement of supercritical CO2 critical flow: Effects of L/D and surface roughness

International Nuclear Information System (INIS)

Mignot, Guillaume P.; Anderson, Mark H.; Corradini, Michael L.

2009-01-01

The use of supercritical fluids (SCF) has been proposed for advanced power systems including advanced sodium reactors, since these fluids can provide higher thermal efficiency and reduced system component size. Data characterizing the behavior of SCF during a blowdown or rapid depressurization are essential to validate certain aspects of safety analyses. This paper describes the results of an experiment to measure the critical mass flux for numerous stagnation thermodynamic conditions, geometry and outlet tube roughness. It was found that a 1D homogeneous equilibrium model (HEM) was capable of relatively good (less than 10% error) prediction of the test data.

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

Resolution of thermal-hydraulic safety and licensing issues for the system 80+trademark design

International Nuclear Information System (INIS)

Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

1995-01-01

The System 80+ trademark Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC's new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

International Nuclear Information System (INIS)

Baratta, A.J.

1997-01-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

Energy Technology Data Exchange (ETDEWEB)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

Energy Technology Data Exchange (ETDEWEB)

Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

1995-09-01

The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

Energy Technology Data Exchange (ETDEWEB)

Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

2014-05-15

For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

2016-10-15

A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

Energy Technology Data Exchange (ETDEWEB)

Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

2014-12-15

The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

Experimental studies on the thermal stratification and its influence on BLEVEs

Energy Technology Data Exchange (ETDEWEB)

Lin, Wensheng; Gong, Yanwu; Gao, Ting; Gu, Anzhong; Lu, Xuesheng [Institute of Refrigeration and Cryogenics, Shanghai Jiao Tong University, Shanghai 200240 (China)

2010-10-15

The thermal stratification of Liquefied Petroleum Gas (LPG) and its effect on the occurrence of the boiling liquid expanding vapor explosion (BLEVE) have been investigated experimentally. Stratifications in liquid and vapor occur when the LPG tank is heated. The degree of the liquid stratification {beta} increases with an increasing heat flux and decreasing filling ratio. The effect of stratification on the BLEVE has been examined with depressurization tests of LPG. The results show that the pressure recovery for the stratified LPG ({beta} = 1.4) upon sudden depressurization is much lower than that for the isothermal LPG ({beta} = 1). It can be concluded that the liquid stratification decreases the liquid energy and the occurrence of the BLEVE. (author)

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

2017-08-15

An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Directory of Open Access Journals (Sweden)

Takeshi Takeda

2017-08-01

Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

Accident loads for a VVER-440/213 containment

Energy Technology Data Exchange (ETDEWEB)

Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

1995-08-01

Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

PKL-tests, test series IIB (end of blowdown). Vol. 2

International Nuclear Information System (INIS)

Umminger, K.; Mandl, R.; Nopper, H.; Siemens AG Unternehmensbereich KWU, Erlangen

1987-01-01

As part of the federally subsidized research project 1500 287/A0, the system behavior of a 1300 MWe pressurized water reactor (PWR) was investigated during the depressurization phase (end-of-blowdown, EOB), as well as during the refill and reflood phases of a loss of coolant accident involving a large break in the reactor coolant loop. Appropriate modifications to the system and supplementary instrumentation have made it possible to simulate the EOB (as of 26 bar), the refill phase and reflood phase in sequence. This report includes a detailed description of the instrumentation and the data acquisition system used in Test Series PKL IIB. (orig.) With 6 refs., 2 tabs., 60 figs [de

Investigation of the Methane Hydrate Formation by Cavitation Jet

Science.gov (United States)

Morita, H.; Nagao, J.

2015-12-01

Methane hydrate (hereafter called "MH") is crystalline solid compound consisting of hydrogen-bonded water molecules forming cages and methane gas molecules enclosed in the cage. When using MH as an energy resource, MH is dissociated to methane gas and water and collect only the methane gas. The optimum MH production method was the "depressurization method". Here, the production of MH means dissociating MH in the geologic layers and collecting the resultant methane gas by production systems. In the production of MH by depressurization method, MH regeneration was consider to important problem for the flow assurance of MH production system. Therefore, it is necessary to clarify the effect of flow phenomena in the pipeline on hydrate regeneration. Cavitation is one of the flow phenomena which was considered a cause of MH regeneration. Large quantity of microbubbles are produced by cavitation in a moment, therefore, it is considered to promote MH formation. In order to verify the possible of MH regeneration by cavitation, it is necessary to detailed understanding the condition of MH formation by cavitation. As a part of a Japanese National hydrate research program (MH21, funded by METI), we performed a study on MH formation using by cavitation. The primary objective of this study is to demonstrate the formation MH by using cavitation in the various temperature and pressure condition, and to clarify the condition of MH formation by using observation results.

Analysis of Total Loss of Feedwater for APR1400 using SPACE

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Min; Park, Seok Jeong; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon [KEPCO Engineering and Construction, Deajeon (Korea, Republic of)

2016-10-15

The Total Loss of FeedWater (TLOFW) event is an accident that main feedwater and auxiliary feedwater of secondary side are not supplied to steam generators. APR1400 uses the Safety Depressurization and Vent System (SDVS) for the F and B operation and SDVS is designed to perform the rapid depressurization function of Reactor Coolant System (RCS) through the remote manual operation when TLOFW is occurred. If RCS pressure falls below a Safety Injection Pump (SIP) working pressure, it can be possible to start the F and B operation which injects SIP flow to RCS and releases the RCS vapor and two-phase flow through Pilot Operated Safety Relief Valves (POSRVs) by opening the POSRVs, and then it can be possible to remove the decay heat. The design requirement of SDVS is that the core water level should be maintained at higher than 2 feet from the top of active core during the F and B operation. The TLOFW analysis was carried out to evaluate the capability of decay heat removal for APR1400 using newly developed SPACE code. The analysis results show that the F and B operation with 2 POSRVs and 2 SIPs and the F and B operation with 4 POSRVs and 4 SIPs meet the SDVS design requirement for the fuel cladding temperature. The comparison with RELAP5 shows good agreement and it validates the applicability of SPACE code for this type of accident analysis.

Test methods for determining asphaltene stability in crude oils

Energy Technology Data Exchange (ETDEWEB)

Asomaning, S. [Baker Petrolite, Sugar Land, TX (United States)

2001-07-01

The development of test methods for the determination of the stability of asphaltenes in crude oils was rendered necessary, due to the high cost of remediating asphaltene deposition in harsh production environments, namely the underwater systems in offshore deepwater. The Oliensis Spot Test, two saturates, aromatics, resins and asphaltenes (SARA)-based screens (the Colloidal Instability Index and Asphaltene-Resin ratio), a solvent titration method with near infrared radiation (NIR) solids detection, and live oil depressurization were used for the purposes of this study, to predict the stability of asphaltenes in crude oils with different API gravity. A complete description of the test methods was provided, and the experimental data obtained as a result was presented. Correlation with data on the deposition histories of the oils was used to validate the experimental stability data. The author discussed the effectiveness of the different tests for the prediction of the stability of asphaltenes in crude oils. The prediction of a crude oil's propensity towards asphaltene precipitation was more accurate with the Colloidal Instability Index and the solvent titration method. Live oil depressurization proved to be very effective for the prediction of the stability of asphaltenes for light oils, where most stability tests fail. tabs., 31 figs.

Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

International Nuclear Information System (INIS)

Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

2014-01-01

Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

Thermal-hydraulic behavior on break simulation of steam generator U-tube

International Nuclear Information System (INIS)

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

1995-01-01

The thermal-hydraulic behavior depending on the break simulation in a steam generator U-tube was investigated and identified the code predictability on plant responses during SGTR accident. The calculated results were compared and assessed with LSTF SB-SG-06 test data. The RELAP5/MOD3.1 code well predicted the sequence of events and the significant phenomena, such as the asymmetric loop behavior, the RCS cooldown and heat transfer by the natural circulation, and system depressurization, even though there were some differences from the experimental data. The break flowrate was found to be sensitive to the break model and affected the system behavior

Alternate method of decay-heat removal in a C-E plant following a SBLOCA

International Nuclear Information System (INIS)

Boyack, B.E.

1986-01-01

The use of an atmospheric steam-dump procedure to cool and depressurize a Combustion-Engineering plant, Calvert Cliffs-1, following small-break loss-of-coolant accidents with failure of the high-pressure injection system to operate has been investigated. The procedure was effective in depressuizing the primary to the low-pressure injection system operating pressure and design temperature using water supplies from the safety-grade condensate water storage tank only. The procedure was found to be effective even if additional failures occurred. Specifically, low-pressure injection conditions were attained if only a single atmospheric dump valve was available or if the safety-injection tanks (accumulators) were not available

Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Conklin, J.C.

1986-08-01

The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

International Nuclear Information System (INIS)

Henninger, R.J.; Boyack, B.E.

1986-01-01

A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

Hydrodynamic calculation of a filter washing in liquids type used in containment venting systems

International Nuclear Information System (INIS)

Reyes G, A. A.; Sainz M, E.; Ortiz V, J.

2015-09-01

From the nuclear accident of Chernobyl, the European nuclear power plants have chosen to install filters on the venting pipes of the containment, whose function is to help to mitigate the consequences of a severe accident, by controlled depressurization of the containment passively through a filtered venting of the containment system. These systems are designed to relieve the internal pressure of the containment by means of the deliberate opening of pressure relief devices, either a valve or rupture disc during a severe accident and be channeled to the filter unit. In this paper the hydraulic response of a filter system of gases washing by liquid is evaluated, due to this information is necessary to estimate the effect that has the pressure increase of the contention on the discharge capacity of the venting pipes. By simulation of computational of fluid dynamics with the programs: CAELINUX-2014 and OpenFOAM, the hydrodynamic characteristics of the Multi Venturi System for gases washing from the containment, which could be included in the general model of the venting pipe, were obtained. Representative models of the Venturi tubes of each concentric area that forming the washing system were generated; and using parametric calculations the average mass flow rate established through each venturi, depending on its size and depth in which it is located inside the tank was estimated. Also, the pressure and mass flow rate required to activate each concentric area depending on the pressure and mass load from the containment were calculated, to estimate the maximum flow that is established through the filter. Finally, the velocity profiles and the characteristic pressure at which each area operates as well as the pressure drop of local and global discharge also were calculated. (Author)

Experiment data report for semiscale Mod-1 test S-04-2 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-2 of the Semiscale Mod-1 Baseline ECC test series. This test is among Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-2 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using emergency core coolant injection parameters based on downcomer volume scaling. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such sime that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-04-2 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent

The LOFA analysis of fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Z.-C.; Xie, H.

2014-01-01

The fusion-fission hybrid energy 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, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (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.

Experimental investigation of methane release from hydrate formation in sandstone through both hydrate dissociation and CO{sub 2} sequestration

Energy Technology Data Exchange (ETDEWEB)

Husebo, J.; Graue, A.; Kvamme, B. [Bergen Univ., Bergen (Norway). Dept. of Physics and Technology; Stevens, J.; Howard, J.J. [ConocoPhillips, Ponca City, OK (United States); Baldwin, B.A. [Green Country Petrophysics LLC, Dewey, OK (United States)

2008-07-01

Large amounts of natural gas trapped in hydrate reservoirs are found in Arctic regions and in deep offshore locations around the world. Natural gas production from hydrate deposits offer significant potential for future energy needs. However, research is needed in order to propose potential production schemes for natural gas hydrates. Natural gas molecules can be freed from hydrate structured cages by depressurization, by heating and by exposing the hydrate to a substance that will form a thermodynamically more stable hydrate structure. This paper provided a comparison of two approaches for releasing methane from methane hydrate in porous sandstone. The study scope covered the dissociation rate of methane hydrate in porous media through depressurization, and also referred to previous work done on producing methane from hydrates in sandstone while sequestering carbon dioxide (CO{sub 2}). The study was conducted in a laboratory setting. The paper discussed the experimental design which included the placing of a pressure- and temperature-controlled sample holder inside the bore of a magnetic resonance imager. The experimental procedures were then outlined, with reference to hydrate formation; carbon dioxide sequestration; hydrate dissociation experiments with constant volume; and hydrate dissociation experiments at constant pressure. The constant volume experiments demonstrated that in order to dissociate a large amount of hydrate, the initial depressurization had to be significantly lower than the hydrate stability pressure. 9 refs., 9 figs.

Analysis and testing of W-DHR system for decay heat removal in the lead-cooled ELSY reactor

International Nuclear Information System (INIS)

Bandini, Giacomino; Meloni, Paride; Polidori, Massimiliano; Gaggini, Piero; Labanti, Valerio; Tarantino, Mariano; Cinotti, Luciano; Presciuttini, Leonardo

2009-01-01

An innovative LFR system that complies with GEN IV goals is under design in the frame of ELSY European project. ELSY is a lead-cooled pool-type reactor of about 1500 MW thermal power which normally relies on the secondary system for decay heat removal. Since the secondary system is not safety-grade and must be fully depressurized in case of detection of a steam generator tube rupture, an independent and much reliable decay heat removal (DHR) system is foreseen on the primary side. Owing to the limited capability of the Reactor Vessel Air Cooling System (RVACS) in this large power reactor, additional safety-grade loops equipped with coolers immersed in the primary coolant are necessary for an efficient removal of decay heat. Some of these loops (W-DHR) are of innovative design and may operate with water at atmospheric pressure. In the frame of the ICE program to be performed on the integral facility CIRCE at ENEA/Brasimone research centre within the EUROTRANS European project, integral circulation experiments with core heat transport and heat removal by steam generator will be conducted in a reactor pool-type configuration. Taking advantage from this experimental program, a mock-up of W-DHR heat exchanger will be tested in order to investigate its functional behavior for decay heat removal. Some pre-test calculations of W-DHR heat exchanger operation in CIRCE have been performed with the RELAP5 thermal-hydraulic code in order to support the heat exchanger design and test conduct. In this paper the experimental activity to be conducted in CIRCE and main results from W-DHR pre-test calculations are presented, along with a preliminary investigation of the W-DHR system efficiency in ELSY configuration. (author)

Hydro-bio-geomechanical properties of hydrate-bearing sediments from Nankai Trough

Science.gov (United States)

Santamarina, J.C.; Dai, Shifeng; Terzariol, M.; Jang, Jeonghwan; Waite, William F.; Winters, William J.; Nagao, J.; Yoneda, J.; Konno, Y.; Fujii, T.; Suzuki, K.

2015-01-01

Natural hydrate-bearing sediments from the Nankai Trough, offshore Japan, were studied using the Pressure Core Characterization Tools (PCCTs) to obtain geomechanical, hydrological, electrical, and biological properties under in situ pressure, temperature, and restored effective stress conditions. Measurement results, combined with index-property data and analytical physics-based models, provide unique insight into hydrate-bearing sediments in situ. Tested cores contain some silty-sands, but are predominantly sandy- and clayey-silts. Hydrate saturations Sh range from 0.15 to 0.74, with significant concentrations in the silty-sands. Wave velocity and flexible-wall permeameter measurements on never-depressurized pressure-core sediments suggest hydrates in the coarser-grained zones, the silty-sands where Sh exceeds 0.4, contribute to soil-skeletal stability and are load-bearing. In the sandy- and clayey-silts, where Sh < 0.4, the state of effective stress and stress history are significant factors determining sediment stiffness. Controlled depressurization tests show that hydrate dissociation occurs too quickly to maintain thermodynamic equilibrium, and pressure–temperature conditions track the hydrate stability boundary in pure-water, rather than that in seawater, in spite of both the in situ pore water and the water used to maintain specimen pore pressure prior to dissociation being saline. Hydrate dissociation accompanied with fines migration caused up to 2.4% vertical strain contraction. The first-ever direct shear measurements on never-depressurized pressure-core specimens show hydrate-bearing sediments have higher sediment strength and peak friction angle than post-dissociation sediments, but the residual friction angle remains the same in both cases. Permeability measurements made before and after hydrate dissociation demonstrate that water permeability increases after dissociation, but the gain is limited by the transition from hydrate saturation

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

International Nuclear Information System (INIS)

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

1977-02-01

Recorded test data are presented for Test S-05-1 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-05-1 was conducted from initial conditions of 2263 psia and 544 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 vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

Analysis of SBLOCA branch cold of a PWR without HPSI through the ISA methodology; Analisis de SBLOCA en Rama Fria de un PWR sin HPSI a traves de la Metodologia ISA

Energy Technology Data Exchange (ETDEWEB)

Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayora, J.

2012-07-01

The objective of the study is the analysis of the appropriateness of the firing of the pumps of the primary and the time available for the beginning of the manual depressurization through steam generators.

Measurement of iodine released in a blowdown accident in the HTR-Modul. Final report on flow tests

International Nuclear Information System (INIS)

Zentis, A.

1993-01-01

A passive measuring device has been designed which consists of several filter cartridges of differnt length, and which is placed into the depressurization channel of the reactor. The dependence of the rate of flow through the filter on the flow rate in the depressurization channel must be known in order to be able to derive from the radioactivity deposited and measured in the filters a value indicating the total amount of iodine released. The report explains the basic principles of design of the instrument and of the experiments, and gives an interpretation of results of the flow tests in the AVA (aerodynamic testing facility) at Interatom. These flow tests have shown that it is feasible to determine the order of magnitude of iodine emissions with the given method and instrument. (orig./HP) [de

Effects of natural and forced basement ventilation on radon levels in single-family dwellings. Final report, May 90-Aug 91

International Nuclear Information System (INIS)

Cavallo, A.; Gadsby, K.; Reddy, T.A.

1992-06-01

The report gives, for the first time, results of an extensive study of the effect of ventilation on radon concentrations and radon entry rate in a single-family dwelling. Measurements of radon concentrations, building dynamics, and environmental parameters made in Princeton University research houses over several seasons and under different building operating conditions show the functional dependence of radon entry rate on basement depressurization. The work clarifies the role of natural ventilation in reducing indoor radon concentrations. The work shows conclusively that natural ventilation can decrease radon levels two ways: (1) by simple dilution, and (2) by providing a pressure break (defined as any opening in the building shell that reduces the outdoor/indoor differential pressure). This reduces building depressurization and thus the amount of radon-contaminated soil gas that is drawn into the building

Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Katanishi, Shoji; Kunitomi, Kazuhiko

2003-01-01

Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Rapid Evaporation of microbubbles

Science.gov (United States)

Gautam, Jitendra; Esmaeeli, Asghar

2008-11-01

When a liquid is heated to a temperature far above its boiling point, it evaporates abruptly. Boiling of liquid at high temperatures can be explosive and destructive, and poses a potential hazard for a host of industrial processes. Explosive boiling may occur if a cold and volatile liquid is brought into contact with a hot and non-volatile liquid, or if a liquid is superheated or depressurized rapidly. Such possibilities are realized, for example, in the depressurization of low boiling point liquefied natural gas (LNG) in the pipelines or storage tanks as a result of a leak. While boiling of highly heated liquids can be destructive at macroscale, the (nearly) instantaneous pace of the process and the release of large amount of kinetic energy make the phenomena extremely attractive at microscale where it is possible to utilize the released energy to derive micromechanical systems. For instance, there is currently a growing interest in micro-explosion of liquid for generation of micro bubbles for actuation purposes. The aim of the current study is to gain a fundamental understanding of the subject using direct numerical simulations. In particular, we seek to investigate the boundary between stable and unstable nucleus growth in terms of the degree of liquid superheat and to compare the dynamics of unstable and stable growth.

A study on the influence diagrams for the application to containment performance analysis

International Nuclear Information System (INIS)

Park, Joon Won

1995-02-01

Influence diagrams have been applied to containment performance analysis of Young-Gwang 3 and 4 in an effort to explicitly display the dependencies between events and to treat operator intervention more generally. This study has been initiated to remove the three major drawbacks of the current event tree methodology: 1) Event tree cannot express dependency between events explicitly. 2) Accident Progression Event Tree (APET) cannot represent entire containment system. 3) It is difficult to consider operator intervention with event tree. To resolve these problems, a new approach, i.e., influence diagrams, are proposed. In the present work, the applicability of the influence diagrams have been demonstrated to YGN 3 and 4 containment performance analysis and an assessment of accident management strategies. To show that the results of the application of influence diagrams are reasonable, results are compared with that of YGN 3 and 4 IPE. Both results are in good agreement. In addition, influence diagrams are used to assess two accident management strategies: 1) RCS depressurization, 2) cavity flooding. Cavity flooding has a favorable effect to late containment failure and basemat melt-through, and depressurization of RCS is good for steam generator tube rupture. However, early containment failure probability is worse in both cases. As a result of the present study, it is shown that influence diagrams can be applied to the containment performance analysis

Coupled analysis of passive safety injection and containment filtered venting for passive decay heat removal - 15140

International Nuclear Information System (INIS)

Kim, S.H.; Ham, J.H.; Jeong, Y.H.; Chang, S.H.

2015-01-01

Lots of interests for the safety of nuclear power plants have risen these days. The safety has to be continuously reviewed and enhanced in nuclear power plants currently operating as well as those designed and constructed in future. After the Fukushima accidents, many additional safety systems which can be applied to nuclear power plants in operation have been proposed. Those include alternating power source such as movable diesel generators and DC batteries in non-safety grade. Also, emergency preparedness for the prevention of a core damage accident was proposed to cope with the extended-SBO (station blackout) by using fire protection systems. In order to prevent the release of radioactive materials, safety systems for preserving the integrity of containment were proposed in two views of cooling and venting containment. Two approaches are effective for mitigating a severe accident. The design concept installing big water tanks besides containment at high level was proposed for various safety functions. One of the functions in the system is to inject the coolant from the elevated tank into a reactor vessel in the case of loss of coolant accident. When the pressure in reactor coolant system is sufficiently low, the coolant can be injected by gravity. If not, the depressurization in reactor vessel would be needed considering the containment pressure. Containment cooling in conventional pressurized water reactors is dependent on containment cooling pumps and sprays. Additional containment cooling systems cannot be simply and easily applied in the current nuclear power plants without major modifications. Therefore, for the operation of passive safety injection system, containment filtered venting system can be adopted for the depressurization of containment. In the design and operation of the passive safety injection system and the containment filtered venting system, main operating points related with open and close pressures in the filtered venting system were

Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

International Nuclear Information System (INIS)

2014-02-01

The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

Simulation of Unique Pressure Changing Steps and Situations in Psa Processes

Science.gov (United States)

Ebner, Armin D.; Mehrotra, Amal; Knox, James C.; LeVan, Douglas; Ritter, James A.

2007-01-01

A more rigorous cyclic adsorption process simulator is being developed for use in the development and understanding of new and existing PSA processes. Unique features of this new version of the simulator that Ritter and co-workers have been developing for the past decade or so include: multiple absorbent layers in each bed, pressure drop in the column, valves for entering and exiting flows and predicting real-time pressurization and depressurization rates, ability to account for choked flow conditions, ability to pressurize and depressurize simultaneously from both ends of the columns, ability to equalize between multiple pairs of columns, ability to equalize simultaneously from both ends of pairs of columns, and ability to handle very large pressure ratios and hence velocities associated with deep vacuum systems. These changes to the simulator now provide for unique opportunities to study the effects of novel pressure changing steps and extreme process conditions on the performance of virtually any commercial or developmental PSA process. This presentation will provide an overview of the cyclic adsorption process simulator equations and algorithms used in the new adaptation. It will focus primarily on the novel pressure changing steps and their effects on the performance of a PSA system that epitomizes the extremes of PSA process design and operation. This PSA process is a sorbent-based atmosphere revitalization (SBAR) system that NASA is developing for new manned exploration vehicles. This SBAR system consists of a 2-bed 3-step 3-layer system that operates between atmospheric pressure and the vacuum of space, evacuates from both ends of the column simultaneously, experiences choked flow conditions during pressure changing steps, and experiences a continuously changing feed composition, as it removes metabolic CO2 and H20 from a closed and fixed volume, i.e., the spacecraft cabin. Important process performance indicators of this SBAR system are size, and the

Safety analyses for an in-pile SCWR fuel qualification test loop

Energy Technology Data Exchange (ETDEWEB)

Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

2014-07-01

As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

Shallow system rejuvenation and magma discharge trends at Piton de la Fournaise volcano (La Réunion Island)

Science.gov (United States)

Coppola, D.; Di Muro, A.; Peltier, A.; Villeneuve, N.; Ferrazzini, V.; Favalli, M.; Bachèlery, P.; Gurioli, L.; Harris, A. J. L.; Moune, S.; Vlastélic, I.; Galle, B.; Arellano, S.; Aiuppa, A.

2017-04-01

Basaltic magma chambers are often characterized by emptying and refilling cycles that influence their evolution in space and time, and the associated eruptive activity. During April 2007, the largest historical eruption of Piton de la Fournaise (Île de La Réunion, France) drained the shallow plumbing system (> 240 ×106 m3) and resulted in collapse of the 1-km-wide summit crater. Following these major events, Piton de la Fournaise entered a seven-year long period of near-continuous deflation interrupted, in June 2014, by a new phase of significant inflation. By integrating multiple datasets (lava discharge rates, deformation, seismicity, gas flux, gas composition, and lava chemistry), we here show that the progressive migration of magma from a deeper (below sea level) storage zone gradually rejuvenated and pressurized the above-sea-level portion of the magmatic system consisting of a vertically-zoned network of relatively small-volume magma pockets. Continuous inflation provoked four small (CO2 enrichment of summit fumaroles, and involving emission of less differentiated lavas, to end with, (iii) three short-lived (∼2 day-long) pulses in lava and gas flux, coupled with arrival of cumulative olivine at the surface and deflation. The activity observed at Piton de la Fournaise in 2014 and 2015 points to a new model of shallow system rejuvenation and discharge, whereby continuous magma supply causes eruptions from increasingly deeper and larger magma storage zones. Downward depressurization continues until unloading of the deepest, least differentiated magma triggers pulses in lava and gas flux, accompanied by rapid contraction of the volcano edifice, that empties the main shallow reservoir and terminates the cycle. Such an unloading process may characterize the evolution of shallow magmatic systems at other persistently active effusive centers.

Calculation of the frequency of excedence in Full Spectrum LOCA by FSR; Calculo de la Frecuencia de excedencia en Full Spectrum LOCA mediante metodologia ISA

Energy Technology Data Exchange (ETDEWEB)

Gomez Magan, J. J.; Queral Salazar, C.; Sanchez Perea, M.

2012-07-01

In this application LOCA sequences was taken into account the uncertainty in the size of rupture and the operator action times as cooling and depressurization through steam generators. The simulations were performed using the tool SCAIS, dynamically coupled with MAAP code.

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Kral, P. [Nuclear Research Inst. Rez (Switzerland)

1995-12-31

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

Analysis of the ANO-2 turbine trip test

International Nuclear Information System (INIS)

McDonald, T.A.; Tessier, J.H.; Senda, Y.; Waterman, M.D.

1983-01-01

The start-up tests performed with the Arkansas Nuclear One-Unit Two (ANO-2) plant provided an opportunity for studying the validity of certain integral systems codes. In particular, the turbine trip from 98.2 percent full power test was investigated with the RELAP5/MOD1 (cycle 18) ode. A detailed plant model was developed and used to understand the test reports. The early depressurization portion of the transient was reproduced; however, the resultant repressurization was not well represented due to uncertainty in the data and plant response. As a result of these computations and detailed analyses of the test data considerable insight was drawn as to the best way to perform and gather data from such integral systems tests for use in code verification studies

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Kral, P [Nuclear Research Inst. Rez (Switzerland)

1996-12-31

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

International Nuclear Information System (INIS)

Kral, P.

1995-01-01

The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

Assessment of SPACE code for multiple failure accident: 1% Cold Leg Break LOCA with HPSI failure at ATLAS Test Facility

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Hyuk; Lee, Seung Wook; Kim, Kyung-Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Design extension conditions (DECs) is a popular key issue after the Fukushima accident. In a viewpoint of the reinforcement of the defense in depth concept, a high-risk multiple failure accident should be reconsidered. The target scenario of ATLAS A5.1 test was LSTF (Large Scale Test Facility) SB-CL-32 test, a 1% SBLOCA with total failure of high pressure safety injection (HPSI) system of emergency core cooling system (ECCS) and secondary side depressurization as the accident management (AM) action, as a counterpart test. As the needs to prepare the DEC accident because of a multiple failure of the present NPPs are emphasized, the capability of SPACE code, just like other system analysis code, is required to expand the DEC area. The objectives of this study is to validate the capability of SPACE code for a DEC scenario, which represents multiple failure accident like as a SBLOCA with HPSI fail. Therefore, the ATLAS A5.1 test scenario was chosen. As the needs to prepare the DEC accident because of a multiple failure of operating NPPs are emphasized, the capability of SPACE code is needed to expand the DEC area. So the capability of SPACE code was validated for one of a DEC scenario. The target scenario was selected as the ATLAS A5.1 test, which is a 1% SBLOCA with total failure of HPSI system of ECCS and secondary side depressurization. Through the sensitivity study on discharge coefficient of break flow, the best fit of integrated mass was found. Using the coefficient, the ATLAS A5.1 test was analyzed using the SPACE code. The major thermal hydraulic parameters such as the system pressure, temperatures were compared with the test and have a good agreement. Through the simulation, it was concluded that the SPACE code can effectively simulate one of multiple failure accidents like as SBLOCA with HPSI failure accident.

Maximum Recoverable Gas from Hydrate Bearing Sediments by Depressurization

KAUST Repository

Terzariol, Marco; Goldsztein, G.; Santamarina, Carlos

2017-01-01

financial analyses; results highlight the need for innovative production strategies in order to make hydrate accumulations an economically-viable energy resource. Horizontal directional drilling and multi-wellpoint seafloor dewatering installations may lead

Overview of the 2006-2008 JOGMEC/NRCan/Aurora Mallik Gas Hydrate Production Test Program

Science.gov (United States)

Yamamoto, K.; Dallimore, S. R.

2008-12-01

During the winters of 2007 and 2008 the Japan Oil, Gas and Metals National Corporation (JOGMEC) and Natural Resources Canada (NRCan), with Aurora Research Institute as the operator, carried out an on-shore gas hydrate production test program at the Mallik site, Mackenzie Delta, Northwest Territories, Canada. The prime objective of the program was to verify the feasibility of depressurization technique by drawing down the formation pressure across a 12m perforated gas hydrate bearing section. This project was the second full scale production test at this site following the 2002 Japex/JNOC/GSC et al Mallik research program in which seven participants organizatinos from five countries undertook a thermal test using hot water circulation Field work in 2007 was devoted to establishing a production test well, installing monitoring devices outside of casing, conducting base line geophysical studies and undertaking a short test to gain practical experience prior to longer term testing planned for 2008 . Hydrate-dissociated gas was produced to surface by depressurization achieved by lowering the fluid level with a dowhole pump. However, the operation was terminated 60 hours after the start of the pumping mainly due to sand production problems. In spite of the short period (12.5 hours of ellapsed pumping time), at least 830m3 of the gas was produced and accumulated in the borehole. Sand screens were installed across the perforated interval at the bottom hole for the 2008 program to overcome operational problems encountered in 2007 and achieve sustainable gas production. Stable bottom hole flowing pressures were successfully achieved during a 6 day test with continuous pump operation. Sustained gas production was achieved with rates between 2000- 4000m3/day and cummulative gas volume in the surface of approximately 13,000m3. Temperature and pressure data measured at the bottom hole and gas and water production rates gave positive evidence for the high efficiency of gas

Rock fracture dynamics research at AECL's Underground Research Laboratory: applications to geological disposal of radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Young, R.P. [Univ. of Toronto, Toronto, ON (Canada); Haycox, J.R. [Applied Seismology Consultants Limited, Shrewsbury, Shropshire (United Kingdom); Martino, J. [Atomic Energy of Canada Limited, Pinawa, MB (Canada)

2011-07-01

Studies of rock fracture dynamics at AECL's Underground Research Laboratory (URL) have helped to provide a fundamental understanding of how crystalline rock responds to stresses induced from excavation, pressurization and temperature changes. The data acquired continue to provide insights into how a facility for the future geological disposal of radioactive waste could be engineered. Research into microseismic (MS), acoustic emission (AE), and ultrasonic velocity measurements has been performed on the full-scale sealed, pressurized, and heated horizontal elliptical tunnel at the Tunnel Sealing Experiment (TSX). The continuous monitoring of the experiment for 8 years provides a unique dataset for the understanding of the medium-term performance of an engineered disposal facility. This paper summarizes the results, interpretations and key findings of the experiment paying particular focus to the heating and cooling/depressurization of the chamber. Initial drilling of the tunnel and bulkheads causes microfracturing around the tunnel, mapped by MS and AEs, and is used as a benchmark for fracturing representing the excavated damaged zone (EDZ). There is no further extension to the volume during pressurization or heating of the tunnel suggesting an increase in crack density and coalescence of cracks rather than extension into unfractured rock. The dominant structure within the seismic cloud has been investigated using a statistical approach applying the three-point method. MS events in the roof exhibit a dominant pattern of sub-horizontal and shallow-dipping well defined planar features, but during cooling and depressurization a 45 degree dip normal to the tunnel axis is observed, which may be caused by movement in the rock-concrete interface due to differential cooling of the bulkhead and host rock. Cooling and depressurization of the TSX have not led to a significant increase in the number of MS or AE events. Ultrasonic results suggest the rock gets even stiffer

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

International Nuclear Information System (INIS)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun; Sun, Xiaodong; Christensen, Richard N.; Oh, Chang H.

2015-01-01

scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed

The complex modelling of various effects of the sub-slab ventilation systems

International Nuclear Information System (INIS)

Svoboda, Z.

2004-01-01

Sub-slab ventilation systems and, in particular, sub-slab depressurization (SSD) systems are among the most efficient radon protective and remedial measures. Numerical modelling can serve as a very powerful tool in the design stage of such systems. The calculations include estimation of the pressure field in the ground under the house with an SSD system and estimation of the radon concentration field. The SSD system also affects the temperature and relative humidity distribution, and therefore those fields should be calculated as well. All the analyses can be carried out applying the simplification of a non-transient steady-state behavior. The numerical solution can be obtained by using the finite difference method or the finite element method. The results of numerical calculation comprise the air pressure field under the building with SSD system, radon concentration field, and temperature and relative humidity fields. The reliability of the numerical models has been verified on six houses with different SSD systems. The results obtained from one house are presented to demonstrate the complete process of verification. The remedial action consisted in the installation of an SSD system in combination with rebuilding of the floors. Soil air temperature, relative humidity, pressure difference and soil air radon concentration were measured continuously. All measurements were carried out for the two modes, i.e. with the SSD system operational or disabled. The first numerical analysis was the calculation of the three-dimensional air pressure field in the whole sub-slab space of the experimental house. The correlation between the calculated and observed values was very good (agreement better than 10%). The calculation of the two-dimensional steady-state temperature and relative humidity field also exhibited a good agreement with the observed values, with differences below 15%. The two-dimensional steady-state field of radon concentrations in the soil under the experimental

SUMMARY OF EPA'S RADON REDUCTION RESEARCH IN SCHOOLS DURING 1989-90

Science.gov (United States)

The report details radon mitigation research in schools conducted by EPA during 1989 and part of 1990. The major objective was to evaluate the potential of active subslab depressurization (ASD) in various geologic and climatic regions. The different geographic regions also pres...

Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

Energy Technology Data Exchange (ETDEWEB)

Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2012-10-15

A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

HANDBOOK: SUB-SLAB DEPRESSURIZATION FOR LOW PERMEABILITY FILL MATERIAL DESIGN AND INSTALLATION OF A HOME RADON REDUCTION SYSTEM

Science.gov (United States)

Radon, a radioactive gas, comes from the natural decay of uranium. It moves to the earth's surface through tiny openings and cracks in soil and rocks. In outdoor air, radon is diluted to such low concentrations that it is usually nothing to worry about. However, radon can accumul...

Comparison between MARCH-3 and MAAP-3 thermal-hydraulic results for a severe accident in a BWR system with MARK-III containment

International Nuclear Information System (INIS)

Barbucci, P.; Guidi, L.; Mariotti, G.

1988-01-01

A comparison between results provided by the Source Term Code Package and by the MAAP-3 code for a PWR with full pressure containment was presented. Thereafter the same two methodologies were used to analyse a severe accident sequence in a typical BWR power plant equipped with a General Electric BWR 6 reactor, rated at 2894 MWt, and a MARK-III type containment. As a reference sequence the TQUV was chosen. This sequence is characterized by a transient (T) with loss of feedwater (Q) and loss of all Emergency Core Cooling Systems, both at high pressure (U) and, after the intervention of the Automatic Depressurization System (ADS), at low pressure (V). After the vessel, failure two basic scenarios for the containment response were analysed: in the first one the pedestal is always dry, in the second one it is fully flooded. Typical limestone/common sand and basaltic concrete compositions were considered. In the following sections the obtained results will be shown with the main purpose to point out the different phenomenological models of the two codes rather than to look for the true plant response to such a severe accident. After the presentation of the most important physical models and of the main assumptions for the analyses (sects. 2 and 4), the comparison will be performed for the in-vessel phase, in section 3, and for the ex-vessel phase, in section 5

A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

International Nuclear Information System (INIS)

Svenson, Ola

1998-01-01

This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

Study on Heat Transfer Characteristics of One Side Heated Vertical Channel Applied as Vessel Cooling System

International Nuclear Information System (INIS)

Kuriyama, Shinji; Takeda, Tetsuaki; Funatani, Shumpei

2014-01-01

The inherent properties of the Very-High-Temperature Reactor facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However; it is still not clear if the VHTR can maintain a passive safe function during the severe accident, or what would be a design criterion to guarantee the VHTR with the high degree of passive safe performances during the accidents. In the Very High Temperature Reactor (VHTR) which is a next generation nuclear reactor system, ceramics and graphite are used as a fuel coating material and a core structural material, respectively. Even if the depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. This is because the thermal capacity of the core is so large. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). This study is to develop the passive cooling system for the VHTR using the vertical channel inserting porous materials. The objective of this study is to investigate heat transfer characteristics of natural convection of a one-side heated vertical channel inserting the porous materials with high porosity. In order to obtain the heat transfer and fluid flow characteristics of a vertical channel inserting porous material, we have also carried out a numerical analysis using the commercial CFD code. From the analytical results obtained in the natural convection cooling, an amount of removed heat enhanced inserting the copper wire. It was found that an amount of removed heat inserting the copper wire (porosity = 0.9972) was about 10% higher than that without the copper wire. This paper describes a thermal performance of the one-side heated vertical channel inserting copper wire with high porosity. (author)

Air pressure distribution and radon entry processes in east Tennessee schools

International Nuclear Information System (INIS)

Sinclair, L.D.; Dudney, C.S.; Wilson, D.L.; Saultz, R.J.

1990-01-01

Many building characteristics have been found to influence radon entry, including building size and configuration, substructure, location of utility supply lines, and design and operation of the heating, ventilation, and air conditioning (HVAC) system. One of the most significant factors is room depressurization resulting from the HVAC system exhausting more than it supplies. This paper represents a preliminary assessment of HVAC characteristics and how they may relate to radon entry. During the summer of 1989, a limited survey was made of air pressure and radon levels in four schools in eastern Tennessee. Short-term samples of radon and pressure were made in all rooms in contact with the soil using alpha scintillation cells and an electronic microanometer, respectively. The pressure difference and radon concentration changes induced by operation of the building ventilation system varied among sites within individual schools

Eden Mills Community Hall energy audit prepared for Eden Mills going carbon neutral

Energy Technology Data Exchange (ETDEWEB)

Lay, R.; Aussant, C. [Enermodal Engineering Ltd., Kitchener, ON (Canada)

2009-04-22

This paper described an energy audit conducted as part of the Eden Mills going carbon neutral project during the spring and summer of 2008. The audit included an inspection of the Eden Mills community hall with a special focus on the building's mechanical system and building envelope. A blower door test was performed to depressurize the building and to measure the airtightness of the building envelope. An energy simulation model was then used to estimate energy use according to the buildings functions and components. Recommendations included the addition of wall insulation, the replacement of some windows, and improved return air ducting and warm air distribution systems. Various new thermostat control systems were also recommended, as well as the use of wood pellets in one of the hall's 2 furnaces. 20 tabs., 28 figs.

Transition from depressurization to long term cooling in AP600 scaled integral test facilities

International Nuclear Information System (INIS)

Bessette, D.E.; Marzo, M. di

1999-01-01

A novel light water reactor design called the AP600 has been proposed by the Westinghouse Electric Corporation. In the evaluation of this plant's behavior during a small break loss of coolant accident (LOCA), the crucial transition to low pressure, long-term cooling is marked by the injection of the gravitationally driven flow from the in-containment refueling water storage tank (IRWST). The onset of this injection is characterized by intermittency in the IRWST flow. This happens at a time when the reactor vessel reaches its minimum inventory. Therefore, it is important to understand and scale the behavior of the integral experimental test facilities during this portion of the transient. The explanation is that the periodic liquid drains and refills of the pressurizer are the reason for the intermittent behavior. The momentum balance for the surge line yields the nondimensional parameter controlling this process. Data from one of the three experimental facilities represent the phenomena well at the prototypical scale. The impact of the intermittent IRWST injection on the safe plant operation is assessed and its implications are successfully resolved. The oscillation is found to result from, in effect, excess water in the primary system and it is not of safety significance. (orig.)

Radon mitigation in schools

International Nuclear Information System (INIS)

Leovic, K.W.; Craig, A.B.; Saum, D.W.

1990-01-01

This article reports on radon mitigation in school buildings. Subslab depressurization (SSD) has been the most successful and widely used radon reduction method in houses. Thus far, it has also substantially reduced radon levels in a number of schools. Schools often have interior footings or thickened slabs that may create barriers for subslab air flow if a SSD system is the mitigation option. Review of foundation plans and subslab air flow testing will help to determine the presence and effect of such barriers. HVAC systems in schools vary considerable and tend to have a greater influence on pressure differentials (and consequently radon levels) than do heating and air-conditioning systems encountered in the radon mitigation of houses. As part of any radon mitigation method, ASHRAE Standard 62-1989 should be consulted to determine if the installed HVAC system is designed and operated to achieve minimum ventilation standards for indoor air quality

Source term estimation for small sized HTRs

International Nuclear Information System (INIS)

Moormann, R.

1992-08-01

Accidents which have to be considered are core heat-up, reactivity transients, water of air ingress and primary circuit depressurization. The main effort of this paper belongs to water/air ingress and depressurization, which requires consideration of fission product plateout under normal operation conditions; for the latter it is clearly shown, that absorption (penetration) mechanisms are much less important than assumed sometimes in the past. Source term estimation procedures for core heat-up events are shortly reviewed; reactivity transients are apparently covered by them. Besides a general literature survey including identification of areas with insufficient knowledge this paper contains some estimations on the thermomechanical behaviour of fission products in water in air ingress accidents. Typical source term examples are also presented. In an appendix, evaluations of the AVR experiments VAMPYR-I and -II with respect to plateout and fission product filter efficiency are outlined and used for a validation step of the new plateout code SPATRA. (orig.)

Underground mining of the lower 163 zone through groundwater drainage at the Eagle Point Mine

International Nuclear Information System (INIS)

Robson, D.M.; Bashir, R.; Thomson, J.; Klemmer, S.; Rigden, A.

2010-01-01

The Eagle Point Mine is part of the Cameco Rabbit Lake Operation. The mine produces uranium ore using the long-hole, vertical and horizontal retreat mining method. The majority of the mine workings are under Wollaston Lake and cementitious grouting is used as one of the water control measures. Historical groundwater table in the mining area was close to ground surface. The Lower 163 Zone encompasses an estimated 4.2 million pounds U_3O_8 geological resource that was not considered feasible to mine due to the expected groundwater flows in the area. Cross-hole testing was conducted to better understand the groundwater flow through various geologic units. A local depressurization test was conducted to assess the potential for lowering the water table. Following testing an active depressurization was conducted to lower the groundwater table below the planned mining areas. This resulted in safe and drier mining conditions and allowed for the successful extraction of the ore body. (author)

On Small Disturbance Ascent Vent Behavior

Science.gov (United States)

Woronowicz, Michael

2015-01-01

As a spacecraft undergoes ascent in a launch vehicle, its ambient pressure environment transitions from one atmosphere to high vacuum in a matter of a few minutes. Venting of internal cavities is necessary to prevent the buildup of pressure differentials across cavity walls. These pressure differentials are often restricted to low levels to prevent violation of container integrity. Such vents usually consist of fixed orifices, ducts, or combinations of both. Duct conductance behavior is fundamentally different from that for orifices in pressure driven flows governing the launch vehicle ascent depressurization environment. Duct conductance is governed by the average pressure across its length, while orifice conductance is dictated by a pressure ratio. Hence, one cannot define a valid equivalent orifice for a given duct across a range of pressure levels. This presentation discusses development of expressions for these two types of vent elements in the limit of small pressure differentials, explores conditions for their validity, and compares their features regarding ascent depressurization performance.

Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Katanishi, Shoji; Kunitomi, Kazuhiko

2003-01-01

Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. (author)

Spray flow-network flow transition of binary Lennard-Jones particle system

KAUST Repository

Inaoka, Hajime

2010-07-01

We simulate gas-liquid flows caused by rapid depressurization using a molecular dynamics model. The model consists of two types of Lennard-Jones particles, which we call liquid particles and gas particles. These two types of particles are distinguished by their mass and strength of interaction: a liquid particle has heavier mass and stronger interaction than a gas particle. By simulations with various initial number densities of these particles, we found that there is a transition from a spray flow to a network flow with an increase of the number density of the liquid particles. At the transition point, the size of the liquid droplets follows a power-law distribution, while it follows an exponential distribution when the number density of the liquid particles is lower than the critical value. The comparison between the transition of the model and that of models of percolation is discussed. The change of the average droplet size with the initial number density of the gas particles is also presented. © 2010 Elsevier B.V. All rights reserved.

Spray flow-network flow transition of binary Lennard-Jones particle system

KAUST Repository

Inaoka, Hajime; Yukawa, Satoshi; Ito, Nobuyasu

2010-01-01

We simulate gas-liquid flows caused by rapid depressurization using a molecular dynamics model. The model consists of two types of Lennard-Jones particles, which we call liquid particles and gas particles. These two types of particles are distinguished by their mass and strength of interaction: a liquid particle has heavier mass and stronger interaction than a gas particle. By simulations with various initial number densities of these particles, we found that there is a transition from a spray flow to a network flow with an increase of the number density of the liquid particles. At the transition point, the size of the liquid droplets follows a power-law distribution, while it follows an exponential distribution when the number density of the liquid particles is lower than the critical value. The comparison between the transition of the model and that of models of percolation is discussed. The change of the average droplet size with the initial number density of the gas particles is also presented. © 2010 Elsevier B.V. All rights reserved.

Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

International Nuclear Information System (INIS)

Nguyen Van Thai; Kieu Ngoc Dung

2015-01-01

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

A framework for the assessment of severe accident management strategies

International Nuclear Information System (INIS)

Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.

1993-09-01

Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed

Study of transient turbine shot without bypass in a BWR

International Nuclear Information System (INIS)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

2015-09-01

The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

Solution of the 6th dynamic AER benchmark using the coupled core DYN3D/ATHLET

International Nuclear Information System (INIS)

Seidel, A.; Kliem, S.

2001-01-01

The 6 th dynamic benchmark is a logical continuation of the work to validate systematically coupled neutron kinetics/thermohydraulics code systems for the estimation of the transient behaviour of WWER type nuclear power plant which was started in the 5 th dynamic benchmark. This benchmark concerns a double ended break of the main steam line (asymmetrical MSLB) in a WWER plant. The core is at the end of first cycle in full power conditions. The asymmetric leak causes a different depressurization of all steam generators. New features in comparison to the 5 th dynamic benchmark were included: asymmetric operation of the feed water system, consideration of incomplete coolant mixing in the reactor vessel, and the definition of a fixed isothermal recriticality temperature for normalising the nuclear data (Authors)

Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie

1998-01-01

Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

Energy Technology Data Exchange (ETDEWEB)

Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2015-09-15

The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

A framework for the assessment of severe accident management strategies

Energy Technology Data Exchange (ETDEWEB)

Kastenberg, W.E. [ed.; Apostolakis, G.; Dhir, V.K. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering] [and others

1993-09-01

Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.

Commercial mitigation techniques used in remediating a 2200 pCi/L public building

International Nuclear Information System (INIS)

Davidson, J.G.

1990-01-01

This paper reports on commercial mitigation techniques used in remediating a 2200 pCi/L public building. In March of 1989 EPA and Pa. DER officials were amazed to discover a school in Pennsylvania with levels in its library of 2200 pCi/L. The library was a 30 year old, three story slab-on-grade structure more like a commercial building than a typical school structure. It had three separate and complex HVAC systems. Initial diagnostics indicated radon levels under the slab at over 80,000 pCi/L. Further investigations revealed major entry routes and a HVAC system terribly out of balance. Remediation consisted of installing a complex sub-slab depressurization system with an exterior commercial fan unit, major entry route sealing, and working closely with a mechanical contractor to bring the HVAC systems back into balance. Initial post remediation testing showed a 99% drop in radon levels. Refinements to the system are still in progress

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

TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

International Nuclear Information System (INIS)

Gallardo, J.; Marquino, W.; Mistreanu, A.; Yang, J.

2015-09-01

The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

Energy Technology Data Exchange (ETDEWEB)

Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

2015-09-15

The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

The establishment of MELCOR/SNAP model of Chinshan nuclear power plant for Ultimate Response Guideline

Energy Technology Data Exchange (ETDEWEB)

Hsu, Wen-Sheng, E-mail: wshsu@ess.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Chiang, Yu, E-mail: s101013702@m101.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Jong-Rong, E-mail: jongrongwang@gmail.com [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Ting-Yi, E-mail: minired1119@gmail.com [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Te-Chuan, E-mail: tcwang@iner.gov.tw [Institute of Nuclear Energy Research Atomic Energy Council, R.O.C., 1000, Wenhua Road Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong, E-mail: jyhtong@cycu.edu.tw [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd, Chung Li 32023, Taiwan, ROC (China); Chen, Shao-Wen, E-mail: chensw@mx.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); and others

2017-01-15

Highlights: • The establishment of a MELCOR/SNAP model of Chinshan (BWR/4). • MELCOR/SNAP model was used to estimate the effectiveness of URG for Chinshan. • The MELCOR results were compared to MAAP, TRACE and PCTRAN. • URG is a new method to prevent a Fukushima-like accident. • The low raw water (150 GPM) can make the cladding temperature below 1088.7 K. - Abstract: After Fukushima Daiichi disaster, the safety analysis of severe accidents became one of the safety concerns in Taiwan. The Emergency Operating Procedure (EOP) cannot cope with a multiple system failure situation under a severe accident since it is a “Symptom-basis” procedure. To deal with that, Taiwan Power Company built up a new strategy for Fukushima-like accident called Ultimate Response Guideline (URG). It is a simple strategy with three main conditions: loss of regular motor driven injection system, loss of all AC power and tsunami/earthquake warning. If two of three happen, the operating procedure will change from EOP to URG and start the main works by following the strategy. There are three main works in URG: controlled-depressurization, line up low pressure injection water and prepare containment venting. In this study, MELCOR2.1 was used to calculate the cases of URG and checked the goal of the strategy that prevents the accident or not. There were three steps in this research. First, a model of Chinshan nuclear power plant (NPP) was built. Second, one was the case with URG and the other was not by using the above MELCOR model. The results were compared to MAAP5.0, TRACE and PCTRAN. Finally, some sensitivity studies of depressurization and water injection rate were done.

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

TRAC analysis of the Crystal River Unit-3 Plant transient of February 26, 1980

International Nuclear Information System (INIS)

Coddington, P.; Willcutt, G.J.E. Jr.

1983-01-01

This paper describes the application of the TRAC-PD2 and TRAC-PF1 codes to analyze the Crystal River transient. The PD2 and PF1 analyses used the three-dimensional and one-dimensional vessel models, respectively. Both calculations predicted the plant depressurization caused by the open PORV and the subsequent repressurization caused by closing the PORV and continuing high-pressure injection flow. Also, natural circulation was calculated in loop B following reestablishment of feedwater to the loop-B steam generator. After system repressurization, the codes calculated that pressure was relieved through the safety valves, and an intermittent flow occurred in loop A because of high-pressure-injection-driven density variations

TRACE Assessment for BWR ATWS Analysis

International Nuclear Information System (INIS)

Cheng, L.Y.; Diamond, D.; Cuadra, Arantxa; Raitses, Gilad; Aronson, Arnold

2010-01-01

A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtained from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depressurization system. The model is not considered complete and recommendations are made on how it should be improved.

Validation of CFD modeling for VGM loss-of-forced-cooling accidents

International Nuclear Information System (INIS)

Wysocki, Aaron; Ahmed, Bobby; Charmeau, Anne; Anghaie, Samim

2009-01-01

Heat transfer and fluid flow in the VGM reactor cavity cooling system (RCCS) was modeled using Computational Fluid Dynamics (CFD). The VGM is a Russian modular-type high temperature helium-cooled reactor. In the reactor cavity, heat is removed from the pressure vessel wall through natural convection and radiative heat transfer to water-cooled vertical pipes lining the outer cavity concrete. The RCCS heat removal capability under normal operation and accident scenarios needs to be assessed. The purpose of the present study is to validate the use of CFD to model heat transfer in the VGM RCCS. Calculations were based on a benchmark problem which defines a two-dimensional temperature distribution on the pressure vessel outer wall for both Depressurized and Pressurized Loss-of-Forced Cooling events. A two-dimensional axisymmetric model was developed to determine the best numerical modeling approach. A grid sensitivity study for the air region showed that a 20 mm mesh size with a boundary layer giving a maximum y+ of 2.0 was optimal. Sensitivity analyses determined that the discrete ordinates radiative model, the k-omega turbulence model, and the ideal gas law gave the best combination for capturing radiation and natural circulation in the air cavity. A maximum RCCS pipe wall temperature of 62degC located 6 m from the top of the cavity was predicted. The model showed good agreement with previous results for both Pressurized and Depressurized Loss-of-Forced-Cooling accidents based on RCCS coolant outlet temperature, relative contributions of radiative and convective heat transfer, and RCCS heat load profiles. (author)

Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

International Nuclear Information System (INIS)

Rimpault, G.; Sugino, K.; Hayashi, H.

2005-01-01

In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

AIR INGRESS ANALYSIS: COMPUTATIONAL FLUID DYNAMIC MODELS

Energy Technology Data Exchange (ETDEWEB)

Chang H. Oh; Eung S. Kim; Richard Schultz; Hans Gougar; David Petti; Hyung S. Kang

2010-08-01

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high temperature reactors (VHTRs). Phenomena Identification and Ranking Studies to date have ranked an air ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the High Temperature Gas Cooled Reactor through the break, possibly causing oxidation of the in-the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of the lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to the release of fission products into the confinement, which could be detrimental to a reactor safety. Computational fluid dynamic model developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional and three-dimensional CFD results for the quantitative assessment of the air ingress phenomena. A portion of results of the density-driven stratified flow in the inlet pipe will be compared with results of the experimental results.

Severe accident management. Optimized guidelines and strategies

International Nuclear Information System (INIS)

Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

2014-01-01

The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

Pressure retarded osmosis as a controlling system for traditional renewables

Science.gov (United States)

Carravetta, Armando; Fecarotta, Oreste; La Rocca, Michele; Martino, Riccardo

2015-04-01

Pressure retarded osmosis (PRO) is a viable but still not diffused form of renewable energy (see Maisonneuve et al., 2015 for a recent literature review). In PRO, water from a low salinity feed solution permeates through a membrane into a pressurized, high salinity draw solution, giving rise to a positive pressure drop; then energy is obtained by depressurizing the permeate through a hydro-turbine and brackish water is discharged. Many technological, environmental and economical aspects are obstacles in the diffusion of PRO, like the vulnerability of the membranes to fouling, the impact of the brackish water on the local marine environment, the high cost of membranes, etc. We are interested in the use of PRO as a combined form of energy with other renewable energy source like solar, wind or mini hydro in water supply networks (WSN). For the wide diffusion of renewables one of the major concerns of commercial power companies is to obtain very stable form of energy to comply with prescriptions of electricity grid operators and with the instant energy demand curve. Renewables are generally very variable form of energy, for the influence of climatic conditions on available power, and of the fluctuation in water demand in WSN. PRO is a very flexible technology where with appropriate turbines and control system power can be varied continuously to compensate for variation of other source of energy. Therefore, PRO is suitable to be used as a balancing system for commercial power system. We will present a simulation of the performance of a PRO used in combination with three different renewables. In the first two scenarios PRO compensate the difference between energy demand and energy production of a solar power plant and hydro power plant in a WSN. In the third scenario PRO is used to compensate daily variation of energy production in a wind power plant. Standard curves of energy production and energy demand for southern Italy are used. In order to control PRO production an

Linear accelerator section alignment in a vacuum chamber

International Nuclear Information System (INIS)

Vengrov, R.M.; Vinogradskij, N.N.; Danil'tsev, E.N.; Iosseliani, D.D.; Kosyak, V.S.; Porubaj, N.I.; Ugarov, S.B.

1989-01-01

Alignment technique for multisectional accelerating structures, that may be used in designing new accelerators for experimental and applied purposes, is described. The accuracy of the alignment of four-chamber resonator sections directly in an accelerator vacuum volume without its depressurization is not less than 100 μm. 8 refs.; 5 figs.; 5 tabs

Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie; Jian Hongbing

1992-01-01

The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

The ultimate emergency measures to secure a NPP under an accidental condition with no designed power or water supply

Energy Technology Data Exchange (ETDEWEB)

Liang, K.S., E-mail: ksliang@alum.mit.edu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong-Chuan Road, Shanghai (China); Chiang, S.C. [Department of Nuclear Safety, Taiwan Power Company, 242 Sec. 3, Roosevelt Road, Taipei 10016, Taiwan (China); Hsu, Y.F.; Young, H.J.; Pei, B.S. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Wang, L.C. [Department of Nuclear Safety, Taiwan Power Company, 242 Sec. 3, Roosevelt Road, Taipei 10016, Taiwan (China)

2012-12-15

Highlights: Black-Right-Pointing-Pointer An ultimate measure to secure core was developed, if power or water supply cannot be restored in time. Black-Right-Pointing-Pointer This ultimate measure was simulated by RELAP5-3D to verify the concept of this emergency plan. Black-Right-Pointing-Pointer Quantification of the required raw water injection rate was performed for NPPS in Taiwan Black-Right-Pointing-Pointer Reactor controlled depressurization within the 1st hour is essential to reduce the required raw water injection rate. Black-Right-Pointing-Pointer For PWR, even heat sink can be developed, RCP seal leak might eventually cause core uncover 10 h after seal leak occurs. - Abstract: In the recent nuclear catastrophe which occurred in Japan on March 11, 2011, several units of Fukushima conventional BWR experienced a total loss of power and water supply triggered by a heavy earthquake and a following Tsunami beyond design basis. In Fukushima accident it was observed that sea water was injected into reactors only after hydrogen explosion took place and it was considered a little too late to prevent core from damage. With regard to this fact, the Taiwan power company develops an ultimate measure to prevent reactor from encountering core damage, if either designed AC power or reactor water supply cannot be restored in time. This ultimate measure was named as DIVing plan, abbreviated from system depressurization, water injection and containment venting. Once any designed AC power or reactor water supply is made available, this DIVing plan will be activated to (1) depressurize reactor first, (2) inject any available water into reactor by any available power supply if this critical status cannot be restored in time, and (3) vent the containment if necessary to maintain containment integrity. In this paper the DIVing plan was simulated by RELAP5-3D to verify the concept of it and also to quantify the required raw water injection rate to prevent core from damage for both

The ultimate emergency measures to secure a NPP under an accidental condition with no designed power or water supply

International Nuclear Information System (INIS)

Liang, K.S.; Chiang, S.C.; Hsu, Y.F.; Young, H.J.; Pei, B.S.; Wang, L.C.

2012-01-01

Highlights: ► An ultimate measure to secure core was developed, if power or water supply cannot be restored in time. ► This ultimate measure was simulated by RELAP5-3D to verify the concept of this emergency plan. ► Quantification of the required raw water injection rate was performed for NPPS in Taiwan ► Reactor controlled depressurization within the 1st hour is essential to reduce the required raw water injection rate. ► For PWR, even heat sink can be developed, RCP seal leak might eventually cause core uncover 10 h after seal leak occurs. - Abstract: In the recent nuclear catastrophe which occurred in Japan on March 11, 2011, several units of Fukushima conventional BWR experienced a total loss of power and water supply triggered by a heavy earthquake and a following Tsunami beyond design basis. In Fukushima accident it was observed that sea water was injected into reactors only after hydrogen explosion took place and it was considered a little too late to prevent core from damage. With regard to this fact, the Taiwan power company develops an ultimate measure to prevent reactor from encountering core damage, if either designed AC power or reactor water supply cannot be restored in time. This ultimate measure was named as DIVing plan, abbreviated from system depressurization, water injection and containment venting. Once any designed AC power or reactor water supply is made available, this DIVing plan will be activated to (1) depressurize reactor first, (2) inject any available water into reactor by any available power supply if this critical status cannot be restored in time, and (3) vent the containment if necessary to maintain containment integrity. In this paper the DIVing plan was simulated by RELAP5-3D to verify the concept of it and also to quantify the required raw water injection rate to prevent core from damage for both PWR and BWR plants in Taiwan, after the loss of passive cooling mechanism. Provided the passive cooling mechanism is lost

MHTGR inherent heat transfer capability

International Nuclear Information System (INIS)

Berkoe, J.M.

1992-01-01

This paper reports on the Commercial Modular High Temperature Gas-Cooled Reactor (MHTGR) which achieves improved reactor safety performance and reliability by utilizing a completely passive natural convection cooling system called the RCCS to remove decay heat in the event that all active cooling systems fail to operate. For the highly improbable condition that the RCCS were to become non-functional following a reactor depressurization event, the plant would be forced to rely upon its inherent thermo-physical characteristics to reject decay heat to the surrounding earth and ambient environment. A computational heat transfer model was created to simulate such a scenario. Plant component temperature histories were computed over a period of 20 days into the event. The results clearly demonstrate the capability of the MHTGR to maintain core integrity and provide substantial lead time for taking corrective measures

Switchable and Tunable Aerodynamic Drag on Cylinders

Science.gov (United States)

Guttag, Mark; Lopéz Jiménez, Francisco; Upadhyaya, Priyank; Kumar, Shanmugam; Reis, Pedro

We report results on the performance of Smart Morphable Surfaces (Smporhs) that can be mounted onto cylindrical structures to actively reduce their aerodynamic drag. Our system comprises of an elastomeric thin shell with a series of carefully designed subsurface cavities that, once depressurized, lead to a dramatic deformation of the surface topography, on demand. Our design is inspired by the morphology of the giant cactus (Carnegiea gigantea) which possesses an array of axial grooves, thought to help reduce aerodynamic drag, thereby enhancing the structural robustness of the plant under wind loading. We perform systematic wind tunnel tests on cylinders covered with our Smorphs and characterize their aerodynamic performance. The switchable and tunable nature of our system offers substantial advantages for aerodynamic performance when compared to static topographies, due to their operation over a wider range of flow conditions.

An experimental investigation of the thermal mixing in a water pool using a simplified I-sparger

International Nuclear Information System (INIS)

Kim, Y. S.; Jun, H. G.; Youn, Y. J.; Park, C. K.; Song, C. H.

2004-01-01

The SDVS (Safety Depressurization and Vent System) in the APR1400 is designed to cope with some DBEs (Design Bases Events) and beyond-DBEs related to overpressurization of the RCS (Reactor Coolant System). When the POSRV (Power Operated Safety Relief Valve) is actuated, steam from the pressurizer is discharged to the IRWST(In-containment Refueling Water Storage Tank) through I-spargers. When injected steam is condensed in the pool, it induces water motions and temperature variations in the pool, which effects on the steam jet condensation, vice versa. The B and C(Blowdown and Condensation) loop is a test facility for the thermal mixing through a steam sparger in a water pool. Thermal mixing tests provide basic understanding of the physics and some insights related to efficient pool mixing, dynamic load, and the IRWST design improvement etc

Radon reduction in crawl-space houses

International Nuclear Information System (INIS)

Osborne, M.C.; Moore, D.G.; Southerlan, R.E.; Brennan, T.; Pyle, B.E.

1989-01-01

This paper gives results of an EPA study of radon-mitigation alternatives for crawl space houses in several houses in Nashville, TN. Application of one of these alternative mitigation options, suction under a polyethylene membrane, has been successful in significantly reducing radon levels in both the crawl space and the house. The large radon concentrations measured under unvented plastic ground covers and the moisture barriers found in many crawl spaces can act as radon-rich reservoirs capable of contaminating a crawl space and house during periods of depressurization. With the exhaust components of the mitigation system in place, radon levels below the plastic decreased by more than 95% under both passive and active suction conditions. Based on the study, the design of a cost-effective subplastic suction passive radon mitigation system for crawl spaces seems promising

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

International Nuclear Information System (INIS)

Roth, P.A.; Schultz, R.R.; Choi, C.J.

1992-07-01

Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests

From insulation contracting to radon mitigation

International Nuclear Information System (INIS)

West, D.R.

1990-01-01

As the definition of house doctor has evolved over the past ten years and the field of energy services has grown more sophisticated, many contractors have expanded the services they offer their clients. This paper presents the story of one insulation contractor who has found a niche in radon testing and mitigation. The EPA now has a national program for the radon mitigator called the Radon Contractor Proficiency Program. The requirements include attending the Radon Technology for Mitigators course, passing an exam, and taking continuing education. In the Midwest, the most popular mitigation technique is the subslab depressurization system. To draw suction from under the slab, the system can take advantage of an existing sump crock or can penetrate the slab. Interior drain tiles collect water to empty into the crock, providing an excellent pathway to draw from. This mitigation process is explained

Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-04-01

Recorded test data are presented for Test S-05-5 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-05-5 was conducted from initial conditions of 2263 psia and 537 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 the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

Direct observation of characteristic dissociation behaviors of hydrate-bearing cores by rapid-scanning X-ray CT imaging

Energy Technology Data Exchange (ETDEWEB)

Ebinuma, T.; Oyama, H.; Utiumi, T.; Nagao, J.; Narita, H. [National Inst. of Advanced Industrial Science and Technology, Toyohiraku, Sapporo (Japan)

2008-07-01

Methane hydrate has significant potential as a new source of energy. Major considerations in developing production methods of methane from hydrates are the fundamental properties of hydrate-bearing sediments, and the dissociation behavior of methane hydrate and the gas and water flow generated by its dissociation in sediments. Marine methane hydrates occur several hundred meters below the sea floor, in a variety of forms. The pore-space filling-type is considered to be the most suited to exploitation, as it is contained within the pore spaces of sandy sediments, and has relatively larger gas permeability compared to other forms. However, shallow sandy sediments are not usually consolidated, and methane hydrate is unstable at normal pressure and temperature. Therefore, common methods are not suitable, and new experimental methods have been developed to study the properties of hydrate-bearing sediment and its dissociation process. This paper presented the results of an experimental study involving the dissociation of artificial methane-hydrate-bearing sediments. The experiment was performed using X-ray computed tomography in order to directly observe dissociation behavior in the sediments and the gas and water flows generated by dissociation. The paper described the depressurization process and presented a schematic diagram of rapid scanning X-ray computed tomography scanner and core holder with tri-axial structure. The experimental apparatus for dissociation of methane hydrate was also illustrated. The thermal stimulation process and hot water injection process were explained. It was concluded that dissociation by depressurization demonstrated that the temperature reduction induced by depressurization depended on the phase equilibrium state of methane hydrate, and that dissociation preferentially occurred at the periphery of the core. This behavior was due to the heat flux from the outside of the core, where the heat flux controlled the dissociation rate. 10 refs

The use of stability indices in predicting asphaltene problems in upstream and downstream oil operations

Energy Technology Data Exchange (ETDEWEB)

Asomaning, S. [Baker Petrolite, Sugar Land, TX (United States)

2003-07-01

A series of test methods have been developed to determine the stability of asphaltenes in crude oils. They were developed due to the high cost of remediating asphaltene deposition in offshore operations. This study described the characteristics of the Oliensis Spot Test, two saturates, aromatics, resins and asphaltenes (SARA)-based screens (the Colloidal Instability Index and Asphaltene-Resin ratio), a solvent titration method with near infrared radiation (NIR) solids detection, and live oil depressurization. Each method is used to predict the stability of asphaltenes in crude oils with different API gravity. A complete description of the test methods was provided along with experimental data. The effectiveness of the different tests in predicting the stability of asphaltenes in crude oils was also assessed. Results indicate that the prediction of a crude oil's tendency towards asphaltene precipitation was more accurate with the Colloidal Instability Index and the solvent titration method. Live oil depressurization proved to be very effective in predicting the stability of asphaltenes for light oils, where most stability tests fail. tabs., figs.

Gas production potential of disperse low-saturation hydrate accumulations in oceanic sediments

International Nuclear Information System (INIS)

Moridis, George J.; Sloan, E. Dendy

2007-01-01

In this paper, we evaluate the gas production potential of disperse, low-saturation (S H H hydrate-bearing sediments subject to depressurization-induced dissociation over a 10-year production period. We investigate the sensitivity of items (a)-(c) to the following hydraulic properties, reservoir conditions, and operational parameters: intrinsic permeability, porosity, pressure, temperature, hydrate saturation, and constant pressure at which the production well is kept. The results of this study indicate that, despite wide variations in the aforementioned parameters (covering the entire spectrum of such deposits), gas production is very limited, never exceeding a few thousand cubic meters of gas during the 10-year production period. Such low production volumes are orders of magnitude below commonly accepted standards of economic viability, and are further burdened with very unfavorable gas-to-water ratios. The unequivocal conclusion from this study is that disperse, low-S H hydrate accumulations in oceanic sediments are not promising targets for gas production by means of depressurization-induced dissociation, and resources for early hydrate exploitation should be focused elsewhere

Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

Energy Technology Data Exchange (ETDEWEB)

Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

2017-12-15

This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

Variable reluctance displacement transducer temperature compensated to 6500F

International Nuclear Information System (INIS)

1975-01-01

In pressurized water reactor tests, compact instruments for accurate measurement of small displacements in a 650 0 F environment are often required. In the case of blowdown tests such as the Loss of Fluid Test (LOFT) or Semiscale computer code development tests, not only is the initial environment water at 650 0 F and 2200 psi but it undergoes a severe transient due to depressurization. Since the LOFT and Semiscale tests are run just for the purpose of obtaining data during the depressurization, instruments used to obtain the data must not give false outputs induced by the change in environment. A LOFT rho v 2 probe and a Semiscale drag disk are described. Each utilizes a variable reluctance transducer (VRT) for indication of the drag-disk location and a torsion bar for drag-disk restoring force. The VRT, in addition to being thermally gain and null offset stable, is fabricated from materials known to be resistant to large nuclear radiation levels and has successfully passed a fast neutron radiation test of 2.7 x 10 17 nvt without failure

Numerical simulation of draining and drying procedure for the ITER Generic Equatorial Port Plug cooling system

International Nuclear Information System (INIS)

Tanchuk, Victor; Grigoriev, Sergey; Lyublin, Boris; Maquet, Philippe; Senik, Konstantin; Pak, Sunil; Udintsev, Victor

2016-01-01

Highlights: • The cooling system of the ITER Generic Equatorial Port Plug (GEPP) is of a complicated combination of horizontal and vertical channels. • The calculation model for the entire GEPP cooling circuit comprising 12 sub-circuits and built up of 2421 finite-volume elements has been developed. • Transient analysis of this model simulating the draining procedure by the KORSAR/B1 code has been performed. • Water in amount of 263 g of initial 531 kg in the GEPP remains in the dead-ends of the DSM and DFW channels in 150 s of draining procedure. • Almost 3 h are required to boil off 263 g of water trapped in the dead-ends. - Abstract: For effective vacuum leak testing all cooling circuits serving the ITER vessel and in-vessel components shall be drained and dried so that after this procedure taking less than 100 h the purge gas passing through a component has water content less than 100 ppm. This process is four-stage, with the first stage using a short blast of compressed nitrogen to blow most of water in the coolant channels out of the circuit. This process is hindered by volumes which trap water due to gravity. To remove the trapped water, it is necessary, first, to heat up the structure by hot and compressed nitrogen, and then water is evaporated by depressurized nitrogen. The cooling system of the ITER Diagnostic Equatorial Port Plugs is of a complicated hydraulic configuration. The system branching might make difficult removal of water from the piping in the scheduled draining mode. The authors have proposed the KORSAR computation code to simulate draining of the GEPP cooling circuit. The numerical simulation performed has made it possible to describe the process dynamics during draining of the entire GEPP cooling circuit and to define the process time, amount and location of residual water and evolution of two-phase flow regime.

Numerical simulation of draining and drying procedure for the ITER Generic Equatorial Port Plug cooling system

Energy Technology Data Exchange (ETDEWEB)

Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Grigoriev, Sergey; Lyublin, Boris [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Maquet, Philippe [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Senik, Konstantin [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Pak, Sunil [National Fusion Research Institute, Daejeon (Korea, Republic of); Udintsev, Victor [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

2016-11-01

Highlights: • The cooling system of the ITER Generic Equatorial Port Plug (GEPP) is of a complicated combination of horizontal and vertical channels. • The calculation model for the entire GEPP cooling circuit comprising 12 sub-circuits and built up of 2421 finite-volume elements has been developed. • Transient analysis of this model simulating the draining procedure by the KORSAR/B1 code has been performed. • Water in amount of 263 g of initial 531 kg in the GEPP remains in the dead-ends of the DSM and DFW channels in 150 s of draining procedure. • Almost 3 h are required to boil off 263 g of water trapped in the dead-ends. - Abstract: For effective vacuum leak testing all cooling circuits serving the ITER vessel and in-vessel components shall be drained and dried so that after this procedure taking less than 100 h the purge gas passing through a component has water content less than 100 ppm. This process is four-stage, with the first stage using a short blast of compressed nitrogen to blow most of water in the coolant channels out of the circuit. This process is hindered by volumes which trap water due to gravity. To remove the trapped water, it is necessary, first, to heat up the structure by hot and compressed nitrogen, and then water is evaporated by depressurized nitrogen. The cooling system of the ITER Diagnostic Equatorial Port Plugs is of a complicated hydraulic configuration. The system branching might make difficult removal of water from the piping in the scheduled draining mode. The authors have proposed the KORSAR computation code to simulate draining of the GEPP cooling circuit. The numerical simulation performed has made it possible to describe the process dynamics during draining of the entire GEPP cooling circuit and to define the process time, amount and location of residual water and evolution of two-phase flow regime.

Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test)

International Nuclear Information System (INIS)

1977-06-01

Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart 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-06-5 was conducted from initial conditions of 2272 psia and 536 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 broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations

A new small HTGR power plant concept with inherently safe features--An engineering and economic challenge

International Nuclear Information System (INIS)

McDonald, C.F.; Sonn, D.L.

1983-01-01

This paper outlines a small nuclear plant concept which is not meant to replace the large nuclear power plants that will continue to be needed by the industrialized nations, but rather recognizes the needs of the smaller energy user, both for special applications in the US and for the developing nations. The small High-Temperature Gas-Cooled Reactor (HTGR), whose introduction will be very dependent on market forces, represents only one approach to meet these needs. The design of a small power plant that could be inherently safer and that might have costs less than those indicated by the traditional reverse-economy-of-scale effect is discussed. Topics considered include power plant economics, the small steam cycle HTGR thermodynamic cycle, the reactor nuclear heat source layout, the reactor heat removal system (main loop cooling, a vessel cooling system with reactor pressurized, vessel cooling system with reactor depressurized), safety considerations, investment risk protection, the technology base, and applications for the small HTGR plant concept

Cost/benefit analysis of adding a feed-and-bleed capability to Combustion Engineering pressurized-water reactors

International Nuclear Information System (INIS)

Gallup, D.R.; Gahan, E.; Cherdack, R.; Skala, G.

1983-08-01

This report presents the results of a cost/benefit analysis for the addition of a feed-and-bleed capability to the San Onofre Nuclear Generating Station, Unit 2, (SONGS 2). Two cases of feed-and-bleed capability were investigated: 1) adding power-operated relief valves (PORVs) to the pressurizer for depressurization and using the present high-pressure safety-injection (HPSI) system for reactor-coolant-system (RCS) inventory make-up and 2) adding an independent single-train feed-and-bleed system. For the first case, it is estimated that the core-melt frequency would be incrementally reduced by 4.0E-6 per year, a factor of 1.3, at a cost of $2.5 M to $4.3 M depending on when the equipment is installed. For the second case, it is estimated that the core-melt frequency would be incrementally reduced by 1.2E-5 per year, a factor of 3, at a cost of $7.0 M to $10.3 M

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

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

Condensation of steam in horizontal pipes: model development and validation

International Nuclear Information System (INIS)

Szijarto, R.

2015-01-01

This thesis submitted to the Swiss Federal Institute of Technology ETH in Zurich presents the development and validation of a model for the condensation of steam in horizontal pipes. Condensation models were introduced and developed particularly for the application in the emergency cooling system of a Gen-III+ boiling water reactor. Such an emergency cooling system consists of slightly inclined horizontal pipes, which are immersed in a cold water tank. The pipes are connected to the reactor pressure vessel. They are responsible for a fast depressurization of the reactor core in the case of accident. Condensation in horizontal pipes was investigated with both one-dimensional system codes (RELAP5) and three-dimensional computational fluid dynamics software (ANSYS FLUENT). The performance of the RELAP5 code was not sufficient for transient condensation processes. Therefore, a mechanistic model was developed and implemented. Four models were tested on the LAOKOON facility, which analysed direct contact condensation in a horizontal duct

[Electrochemical methods of control of iodine contents in drinks].

Science.gov (United States)

Zakharova, E A; Slepchenko, G B; Kolpakova, E Iu

2001-01-01

The simple and express methods of determination of iodide ions (0.01-0.20 mg/decimeter3) in iodine-enriched drinks by potentiometry and inversion voltamperometry were developed. The studies on influencing a storage time hermetically packaged carbonated beverages, a storage time of the depressurized drinks, stuff of ware on the contents of iodine in drinks are held.

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

Analytical scaling relations to evaluate leakage and intrusion in intermittent water supply systems

Science.gov (United States)

Slocum, Alexander H.; Whittle, Andrew J.

2018-01-01

Intermittent water supplies (IWS) deliver piped water to one billion people; this water is often microbially contaminated. Contaminants that accumulate while IWS are depressurized are flushed into customers’ homes when these systems become pressurized. In addition, during the steady-state phase of IWS, contaminants from higher-pressure sources (e.g., sewers) may continue to intrude where pipe pressure is low. To guide the operation and improvement of IWS, this paper proposes an analytic model relating supply pressure, supply duration, leakage, and the volume of intruded, potentially-contaminated, fluids present during flushing and steady-state. The proposed model suggests that increasing the supply duration may improve water quality during the flushing phase, but decrease the subsequent steady-state water quality. As such, regulators and academics should take more care in reporting if water quality samples are taken during flushing or steady-state operational conditions. Pipe leakage increases with increased supply pressure and/or duration. We propose using an equivalent orifice area (EOA) to quantify pipe quality. This provides a more stable metric for regulators and utilities tracking pipe repairs. Finally, we show that the volume of intruded fluid decreases in proportion to reductions in EOA. The proposed relationships are applied to self-reported performance indicators for IWS serving 108 million people described in the IBNET database and in the Benchmarking and Data Book of Water Utilities in India. This application shows that current high-pressure, continuous water supply targets will require extensive EOA reductions. For example, in order to achieve national targets, utilities in India will need to reduce their EOA by a median of at least 90%. PMID:29775462

Analytical scaling relations to evaluate leakage and intrusion in intermittent water supply systems.

Science.gov (United States)

Taylor, David D J; Slocum, Alexander H; Whittle, Andrew J

2018-01-01

Intermittent water supplies (IWS) deliver piped water to one billion people; this water is often microbially contaminated. Contaminants that accumulate while IWS are depressurized are flushed into customers' homes when these systems become pressurized. In addition, during the steady-state phase of IWS, contaminants from higher-pressure sources (e.g., sewers) may continue to intrude where pipe pressure is low. To guide the operation and improvement of IWS, this paper proposes an analytic model relating supply pressure, supply duration, leakage, and the volume of intruded, potentially-contaminated, fluids present during flushing and steady-state. The proposed model suggests that increasing the supply duration may improve water quality during the flushing phase, but decrease the subsequent steady-state water quality. As such, regulators and academics should take more care in reporting if water quality samples are taken during flushing or steady-state operational conditions. Pipe leakage increases with increased supply pressure and/or duration. We propose using an equivalent orifice area (EOA) to quantify pipe quality. This provides a more stable metric for regulators and utilities tracking pipe repairs. Finally, we show that the volume of intruded fluid decreases in proportion to reductions in EOA. The proposed relationships are applied to self-reported performance indicators for IWS serving 108 million people described in the IBNET database and in the Benchmarking and Data Book of Water Utilities in India. This application shows that current high-pressure, continuous water supply targets will require extensive EOA reductions. For example, in order to achieve national targets, utilities in India will need to reduce their EOA by a median of at least 90%.

Hydrodynamic calculation of a filter washing in liquids type used in containment venting systems; Calculo hidrodinamico de un filtro tipo lavado en liquidos usados en los sistemas de venteo de la contencion

Energy Technology Data Exchange (ETDEWEB)

Reyes G, A. A.; Sainz M, E.; Ortiz V, J., E-mail: alejandroantonioreyess@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

From the nuclear accident of Chernobyl, the European nuclear power plants have chosen to install filters on the venting pipes of the containment, whose function is to help to mitigate the consequences of a severe accident, by controlled depressurization of the containment passively through a filtered venting of the containment system. These systems are designed to relieve the internal pressure of the containment by means of the deliberate opening of pressure relief devices, either a valve or rupture disc during a severe accident and be channeled to the filter unit. In this paper the hydraulic response of a filter system of gases washing by liquid is evaluated, due to this information is necessary to estimate the effect that has the pressure increase of the contention on the discharge capacity of the venting pipes. By simulation of computational of fluid dynamics with the programs: CAELINUX-2014 and OpenFOAM, the hydrodynamic characteristics of the Multi Venturi System for gases washing from the containment, which could be included in the general model of the venting pipe, were obtained. Representative models of the Venturi tubes of each concentric area that forming the washing system were generated; and using parametric calculations the average mass flow rate established through each venturi, depending on its size and depth in which it is located inside the tank was estimated. Also, the pressure and mass flow rate required to activate each concentric area depending on the pressure and mass load from the containment were calculated, to estimate the maximum flow that is established through the filter. Finally, the velocity profiles and the characteristic pressure at which each area operates as well as the pressure drop of local and global discharge also were calculated. (Author)

Absorption cycle commercial refrigerator using wood burning cook stove; Geladeira de absorcao acionada por fogao a lenha

Energy Technology Data Exchange (ETDEWEB)

Pereira, Jose Tomaz Vieira; Martins, Gilberto [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Energia

1991-12-31

The current utilization of wood burning cook stoves in Brazil and the socio-economical profile of their users were surveyed. A traditional heavy-mass wood-burning cook stove was studied as a thermal equipment. Simple changes in the geometry of the combustion chamber were suggested to improve the cooking efficiency. A closed two-phase thermosyphon using water as working fluid was designed, built and connected between the combustion chamber of the cook stove and a depressurized absorption refrigeration system to determine the heat flux and the temperature level. A commercial refrigerator unit, using the absorption cycle, was coupled with the wood stove through the thermosyphon. The overall results of the coupling point to successful country-side applications. (author) 12 refs., 9 figs., 4 tabs.

Sensor for measurement of fuel rod gas pressure during loss-of-fluid-tests

International Nuclear Information System (INIS)

Billeter, T.R.

1979-05-01

Qualification tests have been conducted of a measurement system for determining the pressure of certain fuel rods in the loss-of-fluid-test (LOFT) reactor. Because of physical size (0.35-in. OD by 5.5-in length) and operational characteristics, an eddy current device was selected as the most promising measurement transducer for the application. The sensor must operate at pressure up to 17.2 MPa (2500 psig) and at temperatures up to 800 0 F. During the reactor transient caused by loss of coolant flow, sensor temperature and applied pressure will vary rapidly and significantly. Consequently, qualification tests included subjection of the sensor to rapid depressurization, temperature transients, and blowdowns in an autoclave, as well as to calibrations and various slow temperature cycles

ASBWR

International Nuclear Information System (INIS)

Duncan, J.D.; McCandless, R.J.

1988-01-01

This paper reports on the ASBWR, a natural circulation boiling water reactor (BWR). The simplified concept has great potential to produce reduced capital costs, reduced operating and maintenance costs and short construction schedules. Operations are potentially simpler than current designs and require a small operating and engineering support staff. In addition safety characteristics are considerably enhanced. In an emergency, the reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system. No operator action is required. The concept does not require emergency diesel generators, core cooling pumps or heat removal pumps. This is expected to simplify the plant design, reduce costs and simplify licensing

Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

International Nuclear Information System (INIS)

Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

1980-01-01

A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

Experiment data report for LOFT nonnuclear Test L1-4

International Nuclear Information System (INIS)

Batt, D.L.

1977-07-01

Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279 0 C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response

Costs of radon diagnostics and mitigation in school building

International Nuclear Information System (INIS)

Leovic, K.W.; Rector, H.; Nagda, N.

1992-01-01

To determine the costs of radon diagnostics and active soil depressurization (ASD) system installation in schools, seven radon mitigators with extensive experience in school buildings were surveyed. The cost data were determined by providing the mitigators with two scenarios of open-quotes typicalclose quotes school buildings with elevated radon levels. The questionnaire in schools: (1) Review Construction Plans, (2) Conduct Diagnostic Measurements, (3) Design Mitigation System, (4) Purchase ASD Material, and (5) Install and Checkout ASD System. Based on the results of this survey, it is estimated that the average cost of ASD diagnostics and mitigation in a typical school would be roughly $0.50 per ft 2 . However, these costs would be higher in schools with extensive subslab walls, very poor PFE, and lower in simple schools with very good PFE and no subslab barriers to communication. The variations in costs provided by the mitigators are primarily due to the influences of (1) experience and practices of the mitigation companies, (2) ASD system requirements as perceived by the respondents, and (3) the degree of involvement by the school system in the process

Prediction and comparison of noise levels from ground and elevated flare systems

International Nuclear Information System (INIS)

Obasi, E.

2009-01-01

Flaring is a process to dispose of hydrocarbons during clean-up, emergency shut downs or dispose a small volume waste streams of mixed gasses that cannot easily or safely be separated. This presentation discussed flaring as a noise issue. It focused on flaring noise characterization; flare noise modeling; flare sound power levels; and flare sound pressure level comparison at a distance of 1.5 km. The presentation included a photograph of flaring at a gas plant in Nigeria. The presentation listed some of the potential health effects associated with long term exposure to excessive noise, such as hearing loss; headaches; stress; fatigue; sleep disturbance; and high blood pressure. Companies flare gas to dispose waste gases in a safe and reliable manner through combustion and to depressurize gas lines during maintenance and emergencies. This presentation also discussed ground and elevated flares; components of flare noise characterization; and key factors affecting flare noise. A model to predict flaring noise was also presented. It demonstrated that at the same gas mass flow rate, the noise level from elevated flare stacks are significantly higher than ground flares. tabs., figs.

Prediction and comparison of noise levels from ground and elevated flare systems

Energy Technology Data Exchange (ETDEWEB)

Obasi, E. [Stantec Consulting Ltd., Surrey, BC (Canada)

2009-07-01

Flaring is a process to dispose of hydrocarbons during clean-up, emergency shut downs or dispose a small volume waste streams of mixed gasses that cannot easily or safely be separated. This presentation discussed flaring as a noise issue. It focused on flaring noise characterization; flare noise modeling; flare sound power levels; and flare sound pressure level comparison at a distance of 1.5 km. The presentation included a photograph of flaring at a gas plant in Nigeria. The presentation listed some of the potential health effects associated with long term exposure to excessive noise, such as hearing loss; headaches; stress; fatigue; sleep disturbance; and high blood pressure. Companies flare gas to dispose waste gases in a safe and reliable manner through combustion and to depressurize gas lines during maintenance and emergencies. This presentation also discussed ground and elevated flares; components of flare noise characterization; and key factors affecting flare noise. A model to predict flaring noise was also presented. It demonstrated that at the same gas mass flow rate, the noise level from elevated flare stacks are significantly higher than ground flares. tabs., figs.

The application of air pressure difference in reducing indoor radon concentration

International Nuclear Information System (INIS)

Leung, J.K.C.; Tso, M.Y.W.

2000-01-01

In densely populated tropical cities like Hong Kong, people usually live and work inside high-rise buildings. And because of the hot and humid climate, air conditioning systems are used throughout the year, particularly in commercial buildings. Previous territory-wide surveys have shown that over 10% of commercial buildings in Hong Kong have indoor radon concentrations above 200 Bq m -3 . Since the major source of indoor radon in high-rise buildings is the building materials, increasing ventilation and applying radon barriers on wall surfaces seem to be the only ways to reduce the indoor radon concentration. But it was noted that the ventilation rate the many commercial buildings are not efficient enough to remove the radon because of various reasons such as energy saving, lack of maintenance, etc. In this study, radon mitigation was achieved by reducing the rate of radon exhaled from the building materials. A special laboratory, which has the capability of simulating any meteorological conditions that could be faced by high-rise buildings in Hong Kong, was built. The reduction of radon exhalation rate by applying pressure difference and temperature difference across walls was studied in the laboratory. This paper summarizes the results and tactics for applying pressure difference in existing commercial buildings. A new technique of reducing radon exhalation rate in new buildings by depressurizing the interior of walls was also developed. Tunnels can be embedded in the concrete walls of new buildings during construction. By using simple vacuum pumps, radon exhalation rate from the walls can be reduced significantly by depressurizing the tunnels. The feasibility and applicability of the technique is presented in this paper. (author)

Numerical studies of gas production from several CH4 hydrate zones at the Mallik site, Mackenzie Delta, Canada

Science.gov (United States)

Moridis, G.J.; Collett, T.S.; Dallimore, S.R.; Satoh, T.; Hancock, S.; Weatherill, B.

2004-01-01

The Mallik site represents an onshore permafrost-associated gas hydrate accumulation in the Mackenzie Delta, Northwest Territories, Canada. A gas hydrate research well was drilled at the site in 1998. The objective of this study is the analysis of various gas production scenarios from five methane hydrate-bearing zones at the Mallik site. In Zone #1, numerical simulations using the EOSHYDR2 model indicated that gas production from hydrates at the Mallik site was possible by depressurizing a thin free gas zone at the base of the hydrate stability field. Horizontal wells appeared to have a slight advantage over vertical wells, while multiwell systems involving a combination of depressurization and thermal stimulation offered superior performance, especially when a hot noncondensible gas was injected. Zone #2, which involved a gas hydrate layer with an underlying aquifer, could yield significant amounts of gas originating entirely from gas hydrates, the volumes of which increased with the production rate. However, large amounts of water were also produced. Zones #3, #4 and #5 were lithologically isolated gas hydrate-bearing deposits with no underlying zones of mobile gas or water. In these zones, thermal stimulation by circulating hot water in the well was used to induce dissociation. Sensitivity studies indicated that the methane release from the hydrate accumulations increased with the gas hydrate saturation, the initial formation temperature, the temperature of the circulating water in the well, and the formation thermal conductivity. Methane production appears to be less sensitive to the specific heat of the rock and of the hydrate, and to the permeability of the formation. ?? 2004 Published by Elsevier B.V.

Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

International Nuclear Information System (INIS)

Hanson, G.H.

1978-01-01

The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manual activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS

Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture 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-28-3 was conducted from initial conditions of 15621 kPa and 555 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Experiment data of 200% recirculation pump discharge line break integral test run 961 with HPCS failure at ROSA-III and comparison with results of suction line break tests

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Yonomoto, Taisuke; Shiba, Masayoshi

1984-03-01

This report presents the experimental data of RUN 961, a 200% double-ended break test at the recirculation pump discharge line in the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV (Main Steam Isolation Valve) closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The PCT (Peak Cladding Temperature) was 894 K, which was 107 K higher than a 200% pump suction line break test (RUN 926) due to the smaller depressurization rate. The effect of break location on transient LOCA phenomena was clarified by comparing the discharge and suction break tests. The whole core was quenched 71 s after LPCI actuation and the effectiveness of ECCS has been confirmed. (author)

Experiment data report for semiscale Mod-1 test S-06-4 (LOFT counterpart test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.; Coppin, C.E.

1977-12-01

Recorded test data are presented for Test S-06-4 of the Semiscale Mod-1 LOFT counterpart 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-06-4 was conducted from initial conditions of 15,653 kPa and 564 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 100 percent of the maximum peak power density

Experiment data report for semiscale Mod-1 test S-06-1 (LOFT counterpart test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-07-01

Recorded test data are presented for Test S-06-1 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-1 was conducted from initial conditions of 15 568 kPa and 564 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 30% of the maximum peak power density

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

Experiment data report for semiscale Mod-1 Test S-06-2 (LOFT counterpart test)

International Nuclear Information System (INIS)

Patton, M.L. Jr.; Collins, B.L.; Sackett, K.E.

1977-08-01

Recorded test data are presented for Test S-06-2 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-2 was conducted from initial conditions of 15 513 kPa and 563 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 50% of the maximum peak power density

Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

International Nuclear Information System (INIS)

Collins, B.L.; Coppin, C.E.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture 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-28-1 was conducted from initial conditions of 15 767 kPa and 557 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test)

International Nuclear Information System (INIS)

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

1977-10-01

Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture 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-28-2 was conducted from initial conditions of 15 936 kPa and 558 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Esparza, V.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture 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-28-4 was conducted from initial conditions of 15 646 kPa and 557 K 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 broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

An Approach to the Flammability Testing of Aerospace Materials

Science.gov (United States)

Hirsch, David B.

2012-01-01

Presentation reviews: (1) Current approach to evaluation of spacecraft materials flammability (2) The need for and the approach to alternative routes (3) Examples of applications of the approach recommended a) Crew Module splash down b) Crew Module depressurization c) Applicability of NASA's flammability test data to other sample configurations d) Applicability of NASA's ground flammability test data to spacecraft environments

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Westinghouse Small Modular Reactor nuclear steam supply system design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Directory of Open Access Journals (Sweden)

Roman Mukin

2018-04-01

Full Text Available A series of tests dedicated to station blackout (SBO accident scenarios have been recently performed at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility; PKL facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4 of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for

Loads on EPR containment after RPV failure at high pressure; Belastungen des EPR-Containments in Falle eines RDB-Versagens bei hohem Druck

Energy Technology Data Exchange (ETDEWEB)

Jacobs, G.

1995-08-01

As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

Loads on EPR containment after RPV failure at high pressure

International Nuclear Information System (INIS)

Jacobs, G.

1995-01-01

As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

Supercritical impregnation of cinnamaldehyde into polylactic acid as a route to develop antibacterial food packaging materials.

Science.gov (United States)

Villegas, Carolina; Torres, Alejandra; Rios, Mauricio; Rojas, Adrián; Romero, Julio; de Dicastillo, Carol López; Valenzuela, Ximena; Galotto, María José; Guarda, Abel

2017-09-01

Supercritical impregnation was used to incorporate a natural compound with antibacterial activity into biopolymer-based films to develop active food packaging materials. Impregnation tests were carried out under two pressure conditions (9 and 12MPa), and three depressurization rates (0.1, 1 and 10MPamin -1 ) in a high-pressure cell at a constant temperature equal to 40°C. Cinnamaldehyde (Ci), a natural compound with proven antimicrobial activity, was successfully incorporated into poly(lactic acid) films (PLA) using supercritical carbon dioxide (scCO 2 ), with impregnation yields ranging from 8 to 13% w/w. Higher pressure and slower depressurization rate seem to favor the Ci impregnation. The incorporation of Ci improved thermal, structural and mechanical properties of the PLA films. Impregnated films were more flexible, less brittle and more resistant materials than neat PLA films. The tested samples showed strong antibacterial activity against the selected microorganisms. In summary, this study provides an innovative route to the development of antibacterial biodegradable materials, which could be used in a wide range of applications of active food packaging. Copyright © 2017 Elsevier Ltd. All rights reserved.

Studies on air ingress for pebble bed reactors

International Nuclear Information System (INIS)

Moore, R.L.; Oh, C.H.; Merrill, B.J.; Petti, D.A.

2002-01-01

A loss-of-coolant accident (LOCA) has been considered a critical event for helium-cooled pebbled bed reactors. Following helium depressurization, it is anticipated that unless countermeasures are taken air will enter the core through the break and then by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure and graphite pebbles. Thus, without any mitigating features a LOCA will lead to an air ingress event. The INEEL is studying such an event with two well-respected light water reactor transient response codes: RELAP5/ATHENA and MELCOR. To study the degree of graphite oxidation occurring due to an air ingress event, a MELCOR model of a reference pebble bed design was constructed. A modified version of MELCOR developed at INEEL, which includes graphite oxidation capabilities, and molecular diffusion of air into helium was used for these calculations. Results show that the lower reflector graphite consumes all of the oxygen before reaching the core. The results also show a long time delay between the time that the depressurization phase of the accident is over and the time that natural circulation air through the core occurs. (author)

Heat removal tests for pressurized water reactor containment spray by largescale facility

International Nuclear Information System (INIS)

Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

1983-01-01

Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

Development of a test facility for analyzing supercritical fluid blowdown

International Nuclear Information System (INIS)

Roberto, Thiago D.; Alvim, Antonio C.M.

2015-01-01

The generation IV nuclear reactors under development mostly use supercritical fluids as the working fluid because higher temperatures improve the thermal efficiency. Supercritical fluids are used by modern nuclear power plants to achieve thermal efficiencies of around 45%. With water as the supercritical working fluid, these plants operate at a high temperature and pressure. However, experiments on supercritical water are limited by technical and financial difficulties. These difficulties can be overcome by using model fluids, which have more feasible supercritical conditions and exhibit a lower critical pressure and temperature. Experimental research is normally used to determine the conditions under which model fluids represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine model fluids that can represent supercritical fluids in a transient state. This paper presents an application of fractional scale analysis to determine the simulation parameters for a depressurization test facility. Carbon dioxide (CO 2 ) and R134a gas were considered as the model fluids because their critical point conditions are more feasible than those of water. The similarities of water (prototype), CO 2 (model) and R134a (model) for depressurization in a pressure vessel were analyzed. (author)

Some particular aspects of control in nuclear power reactors

International Nuclear Information System (INIS)

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

1964-01-01

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

Thermodynamic correlations for the accident analysis of HTR's

International Nuclear Information System (INIS)

Rehm, W.; Jahn, W.; Finken, R.

1976-12-01

The thermal properties of Helium and for the case of a depressurized primary circuit, various mixtures of primary cooling gas were taken into consideration. The temperature dependence of the correlations for the thermal properties of the graphite components in the core and for the structural materials in the primary circuit are extrapolated about normal operation conditions. Furthermore the correlations for the effective thermal conductivity, the heat transfer and pressure drop are described for pebble bed HTR's. In addition some important heat transfer data of the steam generator are included. With these correlations, for example accident sequences with failure of the afterheat removal systems are discussed for pebble bed HTR's. It is concluded that the transient temperature behaviour demonstrates the inherent safety features of the HTR in extreme accidents. (orig.) [de

RETRAN safety analyses of the nuclear-powered ship Mutsu

International Nuclear Information System (INIS)

Yoshinori, N.; Ishida, T.; Tanaka, Y.; Yoshiaki, F.

1983-01-01

A number of operational transient analyses of the nuclear-powered ship Mutsu have been performed in response to Japanese nuclear safety regulatory concerns. The RETRAN and COBRA-IV computer codes were used to provide a quantitative basis for the safety evaluation of the plant. This evaluation includes a complete loss of load without reactor scram, an excessive load increase incident, and an accidental depressurization of the primary system. The minimum departure from nucleate boiling ratio remained in excess of 1.53 for these three transients. Hence, the integrity of the core was shown to be maintained during these transients. Comparing the transient behaviors with those of land-based pressurized water reactors, the characteristic features of the Mutsu reactor were presented and the safety of the plant under the operational transient conditions was confirmed

Pressurized fluidized bed combustion second-generation system research and development. Technical progress for Phase 2 and Phase 3, October 1, 1997--September 30, 1998

Energy Technology Data Exchange (ETDEWEB)

Robertson, A.; Horazak, D.; Newby, R.; Rehmat, A.; White, J.

1998-10-01

When DOE funds were exhausted in March 1995, all Phase 2 activities were placed on hold. In February 1996 a detailed cost estimate was submitted to the DOE for completing the two remaining Phase 2 Multi Annular Swirl Burner (MASB) topping combustor test burns; in August 1996 release was received from METC to proceed with these tests. The first test (Test Campaign No.3) will be conducted to: (1) test the MASB at proposed demonstration plant full to minimum loading operating conditions; (2) identify the lower oxygen limit of the MASB; and (3) demonstrate natural gas to carbonizer fuel gas switching. The Livingston Phase 3 Pilot Plant was last operated under contract DE-AC21-86MC21023 in September 1995 for seven days in an integrated carbonizer-CPFBC configuration. In May, 1996, the pilot plant was transferred to Contract DE-AC22-95PC95143 to allow testing in support of the High Performance Power Systems (HIPPS) Program. The HIPPS Program required modifications to the pilot plant and the following changes were incorporated: (1) installation of a dense phase transport system for loading pulverized coal into the feed system lock hopper directly from a pneumatic transport truck; (2) removal of the char transfer pipe between the char collecting hopper and the CPFBC to allow carbonizer only operation; (3) installation of a lock hopper directly under the char collecting hopper to facilitate char removal from the process, the hopper vent gases exhaust to the carbonizer baghouse filter and the depressured char is transferred via nitrogen to the CPFBC baghouse for dumping into drums; (4) removal of the carbonizer cyclone and top of bed overflow drain line; all material elutriated from the carbonizer bed will thus be removed by the 22-element Westinghouse ceramic candle filter; (5) replacement of the carbonizer continuous bottom bed drain (screw feeder) with a batch-type drain removal system; and (6) installation of a mass spectrometer that draws sample gas via a steam jacketed

Safety philosophy of the GTHTR300

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Katanishi, Shoji

2003-01-01

In parallel to successful operation of the Japan's first High Temperature Gas-cooed Reactor, HTTR (High Temperature Engineering Test Reactor), JAERI (Japan Atomic Energy Research Institute) started design and development of a high temperature gas cooled reactor with a gas turbine electric generation system, GTHTR300 (Gas Turbine High Temperature Reactor 300), in April 2001. The GTHTR300 is expected to be deployed in 2010s as a safe and economically competitive electric generation system in Japan. Unique safety philosophy is proposed for this system. Severe accidents are defined as any conditions beyond design base accidents, causing core damages with fission product releases to the environment, although all severe accident sequences are very low in probability. The new safety philosophy is to avoid most accidents, and to achieve a probability of severe accidents of 10 -8 /ry that is at least two orders lower than current reactors. Even in the worst event such as double guillotine break of a primary concentric duct, fuel temperature exceeding its failure limit and excessive fuel oxidation by air ingress can be avoided because of inherent safety features and the passive decay heat removal system. Furthermore, double confinement buildings are enough to keep reactor safety in such accidents. Elimination of a leak-tight steel containment vessel is a big economical advantage for this system. Another unique feature is that nearly full-scale worst accident simulation tests can be carried out to obtain licensing before commercial operations because safety assessment by analysis is not usually enough to convince the public and the regulators of trusting this safety concept. In current reactors no accident simulation tests are carried out before commercial operations although inspection and performance tests in normal condition are conducted. This paper describes the safety philosophy together with the outline of the design features of the GTHTR300, and the results of

Annual meeting on nuclear technology 1980. Technical meeting: Shock wave propagation processes

International Nuclear Information System (INIS)

1980-01-01

The papers deal with shock wave propagation processes in LWR-type reactors (licencing procedure) in cases of ruptures in vessels, in pipes and high-pressure valves, in case of loss of coolant accidents with dynamic structure coupling or fluid structure interactions in the reactor core jacket and the fuel rods, as well as with the stresses placed on reactor pressure vessel fittings by depressurization waves. (DG) [de

Fuel elements and fuel element materials. Experimental facilities for fission products lift-off tests

International Nuclear Information System (INIS)

Blanchard, R.J.; Veyrat, J.F.

1978-01-01

One of the hypothetical accidents on the HTGR primary cooling circuits is the failure of a circuit resulting in a depressurization in the primary loops of the reactor. There is a risk of release of fission products in relation to the size of the failure. Experimental facilities for HTGR tests were developed: an in pile helium loop Comedie and an out of pile helium loop

Intentional back flow effects on ruptured steam generator cooldown during a SGTR event for KSNP

International Nuclear Information System (INIS)

Kim, C.W.; Park, S.J.; Choi, C.J.; Seo, J.T.

2004-01-01

For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. One utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability if the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. It is concluded that the

Simulations of nanocrystals under pressure: Combining electronic enthalpy and linear-scaling density-functional theory

Energy Technology Data Exchange (ETDEWEB)

Corsini, Niccolò R. C., E-mail: niccolo.corsini@imperial.ac.uk; Greco, Andrea; Haynes, Peter D. [Department of Physics and Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom); Hine, Nicholas D. M. [Department of Physics and Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom); Cavendish Laboratory, J. J. Thompson Avenue, Cambridge CB3 0HE (United Kingdom); Molteni, Carla [Department of Physics, King' s College London, Strand, London WC2R 2LS (United Kingdom)

2013-08-28

We present an implementation in a linear-scaling density-functional theory code of an electronic enthalpy method, which has been found to be natural and efficient for the ab initio calculation of finite systems under hydrostatic pressure. Based on a definition of the system volume as that enclosed within an electronic density isosurface [M. Cococcioni, F. Mauri, G. Ceder, and N. Marzari, Phys. Rev. Lett.94, 145501 (2005)], it supports both geometry optimizations and molecular dynamics simulations. We introduce an approach for calibrating the parameters defining the volume in the context of geometry optimizations and discuss their significance. Results in good agreement with simulations using explicit solvents are obtained, validating our approach. Size-dependent pressure-induced structural transformations and variations in the energy gap of hydrogenated silicon nanocrystals are investigated, including one comparable in size to recent experiments. A detailed analysis of the polyamorphic transformations reveals three types of amorphous structures and their persistence on depressurization is assessed.

Hydrogen recovery by pressure swing adsorption. [From ammonia purge-gas streams

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

A pressure swing absorption process (PSA) designed to recover H/sub 2/ from ammonia purge-gas streams developed by Bergbarr-Forschung GmbH of West Germany is reviewed. The PSA unit is installed in the process stream after the ammonia absorber unit which washes the ammonia-containing purge gas which consists of NH/sub 3/, H/sub 2/O, CH/sub 4/, Ar, N/sub 2/, and H/sub 2/. Usually 4 absorber units containing carbon molecular sieves make up the PSA unit; however, only one unit is generally used to absorb all components except H/sub 2/ while the other units are being regenerated by depressurization. Economic comparisons of the PSA process with a cryogenic process indicate that for some ammonia plants there may be a 30% saving in fuel gas requirements with the PSA system. The conditions of the purge gas strongly influence which system of recovery is more suitable.

A parametric study of condensation-induced water hammer in nuclear power plants

International Nuclear Information System (INIS)

Shon, Young Uk; Chun, Moon Hyun

1990-01-01

Condensation-induced water hammer (CIWH), which may occur in systems involving steam and water simultaneously, has a series of processes such as formation of water slug, trapping a steam cavity, depressurization due to steam condensation, accelerating slug caused pressure difference over it and final slug impact. These processes are dependent on water flow rate in a pipe, water temperature, water subcooling, steam pressure, size of slug and cavity, and heat transfer coefficient at interface between steam and water. In the present work, the prediction of conditions to initiate water hammer has been made with full scale by applying the open channel flow theory. These conditions are expressed in terms of water flow rate according to changes of steam pressure, water subcooling, and pipe diameter. Under these conditions that induce CIWH, the effect of parameters which influence on slug impact pressure and cavity collapse rate have been studied with full scale. Also, the impact loads that may be applied to piping design were evaluated under various system conditions

Recent advances in severe accident technology - direct containment heating in advanced light water reactors

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

Energy Technology Data Exchange (ETDEWEB)

Chang H. Oh; Eung S. Kim; Hee C. NO; Nam Z. Cho

2011-01-01

The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008–FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

Energy Technology Data Exchange (ETDEWEB)

Sekhri, A.; Baumann, P. [KernkraftwerkLeibstadt AG, 5325 Leibstadt (Switzerland); Wicaksono, D. [Swiss Federal Inst. of Technology Zurich ETH, 8092 Zurich (Switzerland); Miro, R.; Barrachina, T.; Verdu, G. [Inst. for Industrial, Radiophysical and Environmental Safety ISIRYM, Universitat Politecnica de Valencia UPV, Cami de Vera s/n, 46021 Valencia (Spain)

2012-07-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

International Nuclear Information System (INIS)

Sekhri, A.; Baumann, P.; Wicaksono, D.; Miro, R.; Barrachina, T.; Verdu, G.

2012-01-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

Simulation of a main steam line break with steam generator tube rupture using trace

Energy Technology Data Exchange (ETDEWEB)

Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

2012-07-01

A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

Definition of parameters for a test section for the analysis of natural convection and coolant loss in the AP1000 nuclear reactor by similarity laws and fractional scaling analysis

Energy Technology Data Exchange (ETDEWEB)

Cadiz, Luis Felipe S.; Bezerra, Mario Augusto [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lima, Fernando Roberto A., E-mail: falima@cnen.gov.br [Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE/CNEN-PB), Recife, PB (Brazil)

2017-07-01

The present work develops and analyzes the main parameters of a test section for natural convection in case of a failure of the pumping system as much as the loss of coolant in refrigeration accidents. For this realization, a combination of laws of basic similarity and an innovative scale methodology, known as Fractional Scaling Analysis (FSA), was developed. The depressurizing is analyzed when a rupture occurs in one of the primary system piping of the AP1000 nuclear reactor. This reactor is developed by Westinghouse Electric Co., which is a PWR (Pressurized Water Reactor) with an electric power equal to 1000MW. Such a reactor is provided with a passive safety system that promotes considerable improvements in the safety, reliability, protection and reduction of costs of a nuclear power plant. The FSA is based on two concepts: fractional scale and hierarchy. It is used to provide experimental data that generate quantitative evaluation criteria as well as operational parameters in thermal and hydraulic processes of nuclear power plants. The results were analyzed with the use of computational codes. (author)

Cost and effectiveness of radon-resistant features in new school buildings

International Nuclear Information System (INIS)

Craig, A.B.; Leovie, K.W.

1991-01-01

Recent concerns over elevated levels of radon in existing buildings have prompted the design and construction of a number of school buildings that either are radon resistant or incorporate features that facilitate post-construction mitigation if needed. This paper described initial results of a study of several schools with radon-resistant features that were recently constructed in the northeastern U.S. These designs are generally based on experience with radon mitigation in existing houses and schools and radon-resistant new house construction. The study was limited to slab-on-grade schools, where the most common radon-resistant school design is active subslab depressurization (ASD). The additional construction costs for eight schools built with ASD ranged from $3 to $11 per square meter of slab area. The radon contractors who designed these systems have tended to overdesign the radon-reduction systems in the absence of specific written guidance to follow to lessen potential liability in the event of system failure. Design features include detailed sealing of all slab cracks, multiple exhaust stacks, and extensive subslab piping

Impact of Compound Hydrate Dynamics on Phase Boundary Changes

Science.gov (United States)

Osegovic, J. P.; Max, M. D.

2006-12-01

Compound hydrate reactions are affected by the local concentration of hydrate forming materials (HFM). The relationship between HFM composition and the phase boundary is as significant as temperature and pressure. Selective uptake and sequestration of preferred hydrate formers (PF) has wide ranging implications for the state and potential use of natural hydrate formation, including impact on climate. Rising mineralizing fluids of hydrate formers (such as those that occur on Earth and are postulated to exist elsewhere in the solar system) will sequester PF before methane, resulting in a positive relationship between depth and BTU content as ethane and propane are removed before methane. In industrial settings the role of preferred formers can separate gases. When depressurizing gas hydrate to release the stored gas, the hydrate initial composition will set the decomposition phase boundary because the supporting solution takes on the composition of the hydrate phase. In other settings where hydrate is formed, transported, and then dissociated, similar effects can control the process. The behavior of compound hydrate systems can primarily fit into three categories: 1) In classically closed systems, all the material that can form hydrate is isolated, such as in a sealed laboratory vessel. In such systems, formation and decomposition are reversible processes with observed hysteresis related to mass or heat transfer limitations, or the order and magnitude in which individual hydrate forming gases are taken up from the mixture and subsequently released. 2) Kinetically closed systems are exposed to a solution mass flow across a hydrate mass. These systems can have multiple P-T phase boundaries based on the local conditions at each face of the hydrate mass. A portion of hydrate that is exposed to fresh mineralizing solution will contain more preferred hydrate formers than another portion that is exposed to a partially depleted solution. Examples of kinetically closed

TOUGH+HYDRATE v1.2 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

Energy Technology Data Exchange (ETDEWEB)

Moridis, George J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Kowalsky, Michael B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Pruess, Karsten [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2012-08-01

TOUGH+HYDRATE v1.2 is a code for the simulation of the behavior of hydratebearing geologic systems, and represents the second update of the code since its first release [Moridis et al., 2008]. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH4-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy’s law is valid. TOUGH+HYDRATE v1.2 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH4, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is a member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95/2003, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

TOUGH+Hydrate v1.0 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

Energy Technology Data Exchange (ETDEWEB)

Moridis, George; Moridis, George J.; Kowalsky, Michael B.; Pruess, Karsten

2008-03-01

TOUGH+HYDRATE v1.0 is a new code for the simulation of the behavior of hydrate-bearing geologic systems. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH{sub 4}-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy's law is valid. TOUGH+HYDRATE v1.0 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH{sub 4}, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is the first member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

Thermal Hydraulic Assessment for Loss of SDCS Event During the Outage of CANDU Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jonghyun [Gnest, Inc. Taejon (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

2006-07-01

During the outage(overhaul) of the nuclear power plant, there are several operating states other than the full power state, that is 'Hot-Zero Power', 'Depressurized-Cooldown', and 'Partially Drained'. Until now safety assessment has not been done much for this operating state of CANDU type reactor worldwide. For the accuracy and confidence of PSA for the CANDU outage, the safety analysis is necessary. At the first stage, we analyzed the thermal hydraulic characteristics and safety of the postulated event of loss of shutdown cooling system (SDCS) during the partially drained state which is the longest one in the middle of outage period. As an analysis tool, this study uses the best estimate thermal hydraulic code, RELAP5/CANDU which was modified according to the CANDU specific characteristics and based on RELAP5.Mod3.

Creep collapse of thick-walled heat transfer tube subjected to external pressure at high temperature

International Nuclear Information System (INIS)

Ioka, Ikuo; Kaji, Yoshiyuki; Terunuma, Isao; Nekoya, Shin-ichi; Miyamoto, Yoshiaki

1994-09-01

A series of creep collapse tests of thick-walled heat transfer tube were examined experimentally and analytically to confirm an analytical method for creep deformation behavior of a heat transfer tube of an intermediate heat exchanger (IHX) at a depressurization accident of secondary cooling system of HTTR (High Temperature Engineering Test Reactor). The tests were carried out using thick-walled heat transfer tubes made of Hastelloy XR at 950degC in helium gas environment. The predictions of creep collapse time obtained by a general purpose FEM-code ABAQUS were in good agreement with the experimental results. A lot of cracks were observed on the outer surface of the test tubes after the creep collapse. However, the cracks did not pass through the tube wall and, therefore, the leak tightness was maintained regardless of a collapse deformation for all tubes tested. (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)

Fuel Oxidizer Reaction Products (FORP) Contamination of Service Module (SM) and Release of N-nitrosodimethylamine(NDMA)in a Humid Environment from Crew EVA Suits Contaminated with FORP

Science.gov (United States)

Schmidl, William; Mikatarian, Ron; Lam, Chiu-Wing; West, Bil; Buchanan, Vanessa; Dee, Louis; Baker, David; Koontz, Steve

2004-01-01

The Service Module (SM) is an element of the Russian Segment of the International Space Station (ISS). One of the functions of the SM is to provide attitude control for the ISS using thrusters when the U.S. Control Moment Gyros (CMG's) must be desaturated. Prior to an Extravehicular Activity (EVA) on the Russian Segment, the Docking Compartment (DC1) is depressurized, as it is used as an airlock. When the DC1 is depressurized, the CMG's margin of momentum is insufficient and the SM attitude control thrusters need to fire to desaturate the CMG's. SM roll thruster firings induce contamination onto adjacent surfaces with Fuel Oxidizer Reaction Products (FORP). FORP is composed of both volatile and non-volatile components. One of the components of FORP is the potent carcinogen N-nitrosdimethylamine (NDMA). Since the EVA crewmembers often enter the area surrounding the thrusters for tasks on the aft end of the SM and when translating to other areas of the Russian Segment, the presence of FORP is a concern. This paper will discuss FORP contamination of the SM surfaces, the release of NDMA in a humid environment from crew EVA suits, if they happen to be contaminated with FORP, and the toxicological risk associated with the NDMA release.

Space suit glove design with advanced metacarpal phalangeal joints and robotic hand evaluation.

Science.gov (United States)

Southern, Theodore; Roberts, Dustyn P; Moiseev, Nikolay; Ross, Amy; Kim, Joo H

2013-06-01

One area of space suits that is ripe for innovation is the glove. Existing models allow for some fine motor control, but the power grip--the act of grasping a bar--is cumbersome due to high torque requirements at the knuckle or metacarpal phalangeal joint (MCP). This area in particular is also a major source of complaints of pain and injury as reported by astronauts. This paper explores a novel fabrication and patterning technique that allows for more freedom of movement and less pain at this crucial joint in the manned space suit glove. The improvements are evaluated through unmanned testing, manned testing while depressurized in a vacuum glove box, and pressurized testing with a robotic hand. MCP joint flex score improved from 6 to 6.75 (out of 10) in the final glove relative to the baseline glove, and torque required for flexion decreased an average of 17% across all fingers. Qualitative assessments during unpressurized and depressurized manned testing also indicated the final glove was more comfortable than the baseline glove. The quantitative results from both human subject questionnaires and robotic torque evaluation suggest that the final iteration of the glove design enables flexion at the MCP joint with less torque and more comfort than the baseline glove.

MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

International Nuclear Information System (INIS)

Choi, Wonjun; Jeon, Joongoo; Kim, Nam Kyung; Kim, Sung Joong

2017-01-01

The representative example of the severe accident studies using the severe accident code is investigation of effectiveness of developed severe accident management (SAM) strategy considering the positive and adverse effects. In Korea, some numerical studies were performed to investigate the SAM strategy using various severe accident codes. Seo et.al performed validation of RCS depressurization strategy and investigated the effect of severe accident management guidance (SAMG) entry condition under small break loss of coolant accident (SBLOCA) without safety injection (SI), station blackout (SBO), and total loss of feed water (TLOFW) scenarios. The SGTR accident with the sequential mitigation actions according to the flow chart of SAMG was simulated by the MELCOR 1.8.6 code. Three scenariospreventing the RPV failure were investigated in terms of fission product release, hydrogen risk, and the containment pressure. Major conclusions can be summarized as follows: (1) According to the flow chart of SAMG, RPV failure can be prevented depending on the method of RCS depressurization. (2) To reduce the release of fission product during the injecting into SGs, a temporary opening of SDS before the injecting into SGs was suggested. These modified sequences of mitigation actions can reduce the release of fission product and the adverse effect of SDS.

Extended Station Blackout Analyses of an APR1400 with MARS-KS

International Nuclear Information System (INIS)

Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong

2016-01-01

The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage

Soil as a source of indoor 220Rn

International Nuclear Information System (INIS)

Li, Y.; Schery, S.D.; Turk, B.

1992-01-01

Two suggestions for sources of indoor 220Rn (thoron) have appeared in the literature: (1) building materials and outside air, and (2) soil beneath the house. Due to the difficulty of 220Rn measurement and limited data, both suggestions lack sufficient supporting evidence. We have investigated sources of indoor 220Rn in seven occupied houses in northern New Mexico, U.S. A two-filter system was used to measure indoor 220Rn levels continuously, and 220Rn progeny were measured with single filters and specialized alpha-track detectors. The amount of 220Rn entry from soil was curtailed by cutting off soil gas flow to the indoor air with subfloor depressurization mitigation systems. Four of the houses showed significant reductions in 220Rn with mitigation systems on. The average effect for these houses was to reduce indoor 220Rn levels by 70%. The other three houses had no clear reductions but in one of these houses, the mitigation system was not effective for stopping soil gas flow. Our results provide some of the most clear evidence to date supporting soil as an important source of indoor 220Rn

Prediction of the semiscale blowdown heat transfer test S-02-8 (NRC Standard Problem Five)

International Nuclear Information System (INIS)

Fujita, N.; Irani, A.A.; Mecham, D.C.; Sawtelle, G.R.; Moore, K.V.

1976-10-01

Standard Problem Five was the prediction of test S-02-8 in the Semiscale Mod-1 experimental program. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The RELAP4 predictions of Standard Problem 5 were in good overall agreement with the measured hydraulic data. Fortunately, sufficient experience has been gained with the semiscale break configuration and the critical flow models in RELAP4 to accurately predict the break flow and, hence the overall system depressurization. Generally, the hydraulic predictions are quite good in regions where homogeneity existed. Where separation effects occurred, predictions are not as good, and the data oscillations and error bands are larger. A large discrepancy existed among the measured heater rod temperature data as well as between these data and predicted values. Several potential causes for these differences were considered, and several post test analyses were performed in order to evaluate the discrepancies

Extended Station Blackout Analyses of an APR1400 with MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage.

Transient behavior of ASTRID with a gas power conversion system

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr; Mauger, G.; Bensalah, M.; Gauthé, P.

2016-11-15

Highlights: • CATHARE2 transient calculations have been performed for ASTRID with a gas PCS. • The behavior of the reactor is close for gas and for water PCS in case of LOOP. • The gas PCS enables to cool the core for at least 10 h for pressurized transients. • The depressurization of the PCS induces an over-cooling for breaches on low pressure pipes. • The spurious opening of a by-pass line of the turbomachine can be controlled without scram. - Abstract: The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of

Transient behavior of ASTRID with a gas power conversion system

International Nuclear Information System (INIS)

Bertrand, F.; Mauger, G.; Bensalah, M.; Gauthé, P.

2016-01-01

Highlights: • CATHARE2 transient calculations have been performed for ASTRID with a gas PCS. • The behavior of the reactor is close for gas and for water PCS in case of LOOP. • The gas PCS enables to cool the core for at least 10 h for pressurized transients. • The depressurization of the PCS induces an over-cooling for breaches on low pressure pipes. • The spurious opening of a by-pass line of the turbomachine can be controlled without scram. - Abstract: The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of

Multi-Fluid Geo-Energy Systems for Bulk and Thermal Energy Storage and Dispatchable Renewable and Low-Carbon Electricity

Science.gov (United States)

Buscheck, T. A.; Randolph, J.; Saar, M. O.; Hao, Y.; Sun, Y.; Bielicki, J. M.

2014-12-01

Integrating renewable energy sources into electricity grids requires advances in bulk and thermal energy storage technologies, which are currently expensive and have limited capacity. We present an approach that uses the huge fluid and thermal storage capacity of the subsurface to harvest, store, and dispatch energy from subsurface (geothermal) and surface (solar, nuclear, fossil) thermal resources. CO2 captured from fossil-energy systems and N2 separated from air are injected into permeable formations to store pressure, generate artesian flow of brine, and provide additional working fluids. These enable efficient fluid recirculation, heat extraction, and power conversion, while adding operational flexibility. Our approach can also store and dispatch thermal energy, which can be used to levelize concentrating solar power and mitigate variability of wind and solar power. This may allow low-carbon, base-load power to operate at full capacity, with the stored excess energy being available to addresss diurnal and seasonal mismatches between supply and demand. Concentric rings of horizontal injection and production wells are used to create a hydraulic divide to store pressure, CO2, N2, and thermal energy. Such storage can take excess power from the grid and excess thermal energy, and dispatch that energy when it is demanded. The system is pressurized and/or heated when power supply exceeds demand and depressurized when demand exceeds supply. Supercritical CO2 and N2 function as cushion gases to provide enormous pressure-storage capacity. Injecting CO2 and N2 displaces large quantities of brine, reducing the use of fresh water. Geologic CO2 storage is a crucial option for reducing CO2 emissions, but valuable uses for CO2 are needed to justify capture costs. The initial "charging" of our system requires permanently isolating large volumes of CO2 from the atmosphere and thus creates a market for its disposal. Our approach is designed for locations where a permeable

Plugging Effects on Depressurization Time in Dry Storage Containers with Pinhole Breaches

International Nuclear Information System (INIS)

Casella, Andrew M.; LOYALKA, SUDARSHAN K.; Hanson, Brady D.

2006-01-01

As continuation on previous work, we now examine the effect that aerosol deposition may have on plugging pinhole breaches in spent fuel containers. A model is developed considering only diffusive settling

Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

International Nuclear Information System (INIS)

Martin, R.C.

1991-01-01

A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of 137 Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for 85 Kr, 90 Sr, and 110m Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models of fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. As oxygen-enhanced SiC decomposition mechanism is also discussed. (author). 12 refs, 11 figs, 2 tabs

Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

Energy Technology Data Exchange (ETDEWEB)

Logvinov, S.A.; Sitnik, Y.K. [EDO Gidropress, Podolsk (Russian Federation)

1997-12-31

The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

Use of the supercritical fluid technology to prepare efficient nanocomposite foams for environmental protection purpose

OpenAIRE

Urbanczyk, Laetitia; Thomassin, Jean-Michel; Huynen, Isabelle; Alexandre, Michaël; Jérôme, Christine

2009-01-01

This work reports on the preparation of novel nanocomposite foams that are efficient broadband microwave absorbers. Carbon nanotubes are first successfully dispersed into PCL and PMMA by melt blending. Then, foaming is promoted by supercritical CO2 by depressurization. Regular cellular structures are obtained in both cases with cells size around 10-50µm. The electromagnetic interference (EMI) shielding efficiency of these materials are then evaluated and compared to the non-foamed nanocomposi...

A case of death of the driver due to environmental asphyxia by liquid nitrogen leakage in the cabin of the car during a road accident

Science.gov (United States)

Raczkowska, Zuzanna; Samojłowicz, Dorota

2013-01-01

Nitrogen causes environmental asphyxia by displacing oxygen in the air leading to death. The study presents a case of a death of a driver death who was transporting flasks with liquid nitrogen that depressurized during an accident. The mechanism and cause of death were determined based on the result of the autopsy and histopathologic examination. The authors emphasize the relevance of accident scene inspection during establishing the cause of death in similar cases.

Investigations on the propagation of free surface boiling in a vertical superheated liquid column

International Nuclear Information System (INIS)

Das, P.K.; Bhat, G.S.; Arakeri, V.H.

1987-01-01

Some experimental studies on boiling propagation in a suddenly depressurized superheated vertical liquid column are reported. The propagation velocity of this phase change has been measured using an optical method. This velocity is strongly dependent on liquid superheat, liquid purity and test section size. The measured velocities of less than 5 m s -1 are significantly lower than the sonic velocity. Present observations suggest that the dominant mechanism for boiling propagation is convection. (author)

Simulation model of dynamical behaviour of reactor fuel assemblies

International Nuclear Information System (INIS)

Planchard, J.

1994-01-01

This report briefly describes the homogenized dynamical equations of a tube bundle placed in a perfect irrotational fluid, on case of small displacements. This approach can be used to study the mechanical behaviour of fuel assemblies of PWR reactor submitted to earthquake or depressurization blow-down. The numerical calculations require to define the added mass matrix of the fuel assemblies, for which the principle of computation is presented. (author). 14 refs., 4 figs

Unconventional Fossil-Based Fuels. Economic and Environmental Trade-Offs

Science.gov (United States)

2008-01-01

concern with using water for in situ operations is that the water is generally drawn from aquifers (fresh or saline) due to the location of operations...depressurization of freshwater aquifers , changes in groundwater levels, changes in underground water storage or flow due to voidage zones left by bitumen removal...More recently, Sasol converted a CTL facility to accept natural gas from Mozambique. In Qatar , a large GTL facility has recently begun operating, and a

Test plan for In Situ Vitrification Engineering-Scale Test No. 6, EG ampersand G Idaho, Inc., Job Number 318230

International Nuclear Information System (INIS)

1991-03-01

The objectives of the test included the effects of in situ vitrification on containerized sludge contained in a simulated randomly-disposed array. From this arrangement, the test results obtained the following data applicable to Idaho National Engineering Laboratory Large Field Testing: canister burst pressure and temperature, canister depressurization rate, melt encapsulation rate of the canister and the hood area plenum temperatures, pressures, compositional analyses, and flows as affected by gas releases. 10 figs., 1 tab

Investigations on the propagation of free surface boiling in a vertical superheated liquid column

Energy Technology Data Exchange (ETDEWEB)

Das, P.K.; Bhat, G.S.; Arakeri, V.H.

1987-04-01

Some experimental studies on boiling propagation in a suddenly depressurized superheated vertical liquid column are reported. The propagation velocity of this phase change has been measured using an optical method. This velocity is strongly dependent on liquid superheat, liquid purity and test section size. The measured velocities of less than 5 m s/sup -1/ are significantly lower than the sonic velocity. Present observations suggest that the dominant mechanism for boiling propagation is convection.

Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

Energy Technology Data Exchange (ETDEWEB)

Logvinov, S A; Sitnik, Y K [EDO Gidropress, Podolsk (Russian Federation)

1998-12-31

The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

Water spray interaction with air-steam mixtures under containment spray conditions: experimental study in the TOSQAN facility

Energy Technology Data Exchange (ETDEWEB)

Porcheron, E.; Lemaitre, P.; Malet, J.; Nuboer, A.; Brun, P.; Bouilloux, L.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Direction de la Surete des Usines, des laboratoires, des transports et des dechets, Saclay, BP 68 - 91192 Gif-sur-Yvette cedex (France)

2005-07-01

Full text of publication follows: During the course of an hypothetical severe accident in a Pressurized Water Reactor (PWR), hydrogen can be produced by the reactor core oxidation and distributed into the reactor containment according to convection flows and steam wall condensation. In order to assess the risk of detonation generated by a high local hydrogen concentration, hydrogen distribution in the containment has to be known. The TOSQAN experimental program has been created to simulate typical accidental thermal hydraulic flow conditions in the reactor containment. The present work is devoted to study the interaction of a water spray injection used as a mitigation mean in order to reduce containment pressure and to produce a mixing of air, steam and hydrogen induced by spray entrainment and condensation on droplet. In order to have a better understanding of physical phenomena, we need to make a detailed characterization of the spray and the gas. The TOSQAN facility that is highly instrumented with non-intrusive diagnostics consists in a closed cylindrical vessel (7 m{sup 3} volume, 4 m high, 1.5 m i.d.) into which steam is injected. Water droplets size is measured in the vessel by the Interferometric Laser Imaging for Droplet Sizing technique. Droplet velocity is obtained by Particle Image Velocimetry and Laser Doppler Velocimetry, and droplet temperature is measured by global rainbow refractometry. Gas concentration measurements are performed by Spontaneous Raman Scattering. The walls of the vessel are thermostatically controlled by heated oil circulation. Inner spray system that is located on the top of the enclosure on the vertical axis, is composed of a single nozzle producing a full cone water spray. Spray test scenario consists of water spray injection in TOSQAN that is first pressurized with a steam injection (steam injection is stopped before spray injection). Water spray falling into the sump is removed to avoid accumulation and evaporation

Update on radon-mitigation research in schools. Rept. for 1988-Aug 91

International Nuclear Information System (INIS)

Leovic, K.W.; Craig, A.B.; Harris, D.B.

1991-01-01

The paper is an overview of research by EPA's Air and Energy Engineering Research Laboratory (AEERL) on radon mitigation in 47 schools since 1988. The structural and heating, ventilating, and air-conditioning (HVAC) system characteristics of the research schools are presented, along with the mitigation techniques implemented in the schools. Research discussed includes recent and on-going projects in Colorado, Maine, Maryland, Ohio, South Dakota, Tennessee, and Virginia. Initial research focussed on the application of active subslab depressurization (ASD) to school buildings, and recent research has emphasized the ability and limitations of using HVAC systems to reduce radon levels in schools. A goal of future projects is to compare the effectiveness of the two techniques in the same building. Slab-on-grade is the most prevalent substructure in AEERL's research schools and, depending on pressure field extension, ASD systems have been recommended for radon control in many of them. In schools where they have been installed, ASD systems have performed well and are currently being evaluated for long-term performance. The distribution of HVAC system types in these schools is about a third central air handling systems, a third unit ventilators, and a third that do not supply conditioned outdoor air (i.e., fan coil units or radiant heat)

Status of the ETDR design

International Nuclear Information System (INIS)

Poette, C.; Garnier, J.C.; Klein, J.C.; Morin, F.; Tosello, A.; Dor, I.; Bertrand, F.; Every, D.; Coddington, P.

2007-01-01

The Experimental and Technology Demonstrator Reactor (ETDR) will be the first Gas Fast Reactor (GFR) ever built. It is a small power experimental reactor and a necessary step towards an electricity generating prototype GFR. In the fuel development plan, the ETDR is located between the irradiation of samples in Material Testing Reactors and the full demonstration at the GFR prototype scale. After revisiting the main reactor objectives, the paper will give an overview of the progress in various areas like: -) core design including a 3-dimensional core physics analysis of the starting core showing in particular that the control rods safety criteria are satisfied, -) sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding), -) system design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy for de-pressurized accidents. The safety approach is built by reference to a familiar fuel in situations which are estimated to be particularly constraining for the safeguard systems. The main elements of the approach include: -) a proven fuel (with a design providing margins), -) a reliable, performing monitoring and protection system, and -) a reliable, performing decay heat removal system (ensuring primary helium circulation). The system transient analyses will be shared between the European partners using their own system codes. (authors)

Containment design for Indian PHWRs - evolution and future trends

International Nuclear Information System (INIS)

Chatterjee, S.K.; Srinivasan, G.R.; Das, M.; Prakash, P.; Mulgund, S.

1994-01-01

The design of containment systems for PHWRs in India has undergone progressive improvements to enhance their reliability and effectiveness. The state-of-the-art containment design incorporates a double containment structure for minimizing radioactivity release to the environment, a completely passive vapour suppression system with huge suppression pool for limiting pressure build-up during postulated LOCA and various engineered systems for depressurizing the containment and cleaning the containment environment following an accident. The containment related Engineered Safety Features (ESFS) include Reactor Building (RB) coolers, Primary Containment Controlled Discharge (PCCD) system, Primary Containment Filtration and Pump-Back (PCFPB) system and Secondary Containment Filtered Recirculation and Purge (SCFRP) system. Studies indicate that the unique feature of double containment with huge suppression pool at basement and associated ESFs not only ensures near zero ground level release during Design Basis Accident (DBA) conditions, but also provides adequate assurance for containment integrity even in beyond DBA scenario. In this paper, an outline of the containment design evolution in Indian PHWRs is presented and salient features of standardized containment design are highlighted. Important containment related studies are discussed and outstanding safety issues viz. hydrogen generation and management, containment venting, containment over pressure capability, etc. are addressed. (author). 16 refs., 1 tab., 8 figs

Verification of the computer code ATHLET in the framework of the external verification group ATHLET BETHSY test 9.3 - steam generator U-tube rupture with failure of the high pressure injection. Final report; Verifikation des ATHLET-Rechenprogramms im Rahmen der externen Verifikationsgruppe ATHLET BETHSY Test 9.3 - Heizrohrbruch mit Versagen der Hochdruck-Noteinspeisung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Krepper, E.; Schaefer, F. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

1998-08-01

In the framework of the external validation of the thermalhydraulic code ATHLET MOD 1.1 CYCLE D, which is being developed by the GRS, post test analyses of two experiments were done, which were performed at the French integral test facility BETHSY. During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. The safety relevant statement could be reproduced. Deviations, which did not impose the general results, occurred concerning the break mass flow during the depressurization of the damaged steam generator and the description of the failure of the heat transfer to the damaged steam generator. Reasons are hardly to find, because these processes are highly complex. (orig.) [Deutsch] Im Rahmen der externen Validierung des von der Gesellschaft fuer Anlagen- und Reaktorsicherheit entwickelten Stoerfallcodes ATHLET, der in der Version Mod 1.1 Cycle D vorlag, wurden zwei Experimente nachgerechnet und analysiert, die an der franzoesischen Versuchsanlage BETHSY durchgefuehrt wurden. Im Test 9.3 werden die Konsequenzen untersucht, wenn bei einem Heizrohrbruch die Hochdruckeinspeisung sowie die Not-Speisewasserversorgung der Dampferzeuger versagen und nur die Druckspeicher sowie die Niederdruckeinspeisung zur Verfuegung stehen. Als Accident Management Massnahmen wurde die sekundaere Druckentlastung und schliesslich die primaere Entlastung ueber den Druckhalter untersucht. Die Analyse kommt zu dem Ergebnis, dass der Code ATHLET in der Lage ist, dieses komplexe Szenario recht gut zu beschreiben. Die

Numerical simulation of gas hydrate exploitation from subsea reservoirs in the Black Sea

Science.gov (United States)

Janicki, Georg; Schlüter, Stefan; Hennig, Torsten; Deerberg, Görge

2017-04-01

Natural gas (methane) is the most environmental friendly source of fossil energy. When coal is replace by natural gas in power production the emission of carbon dioxide is reduced by 50 %. The vast amount of methane assumed in gas hydrate deposits can help to overcome a shortage of fossil energy resources in the future. To increase their potential for energy applications new technological approaches are being discussed and developed worldwide. Besides technical challenges that have to be overcome climate and safety issues have to be considered before a commercial exploitation of such unconventional reservoirs. The potential of producing natural gas from subsea gas hydrate deposits by various means (e. g. depressurization and/or carbon dioxide injection) is numerically studied in the frame of the German research project »SUGAR - Submarine Gas Hydrate Reservoirs«. In order to simulate the exploitation of hydrate-bearing sediments in the subsea, an in-house simulation model HyReS which is implemented in the general-purpose software COMSOL Multiphysics is used. This tool turned out to be especially suited for the flexible implementation of non-standard correlations concerning heat transfer, fluid flow, hydrate kinetics, and other relevant model data. Partially based on the simulation results, the development of a technical concept and its evaluation are the subject of ongoing investigations, whereby geological and ecological criteria are to be considered. The results illustrate the processes and effects occurring during the gas production from a subsea gas hydrate deposit by depressurization. The simulation results from a case study for a deposit located in the Black Sea reveal that the production of natural gas by simple depressurization is possible but with quite low rates. It can be shown that the hydrate decomposition and thus the gas production strongly depend on the geophysical properties of the reservoir, the mass and heat transport within the reservoir, and

Generation of Microcellular Biodegradable Polycaprolactone Foams in Supercritical Carbon Dioxide

Institute of Scientific and Technical Information of China (English)

Xu Qun; Ren Xian-wen; Chang Yu-ning; Yu Long; Wang Jing-wu

2004-01-01

Present now the application of microcellular polymeric materials in biomedical field is growing rapidly, as that of guided tissue regeneration and cell transplantation. As far as guided tissue regeneration is concerned, porous implants are used as size selective membrane to promote the growth of a special tissue in a healing site. Ideally, the implant should be inherently biocompatible,have well-defined cell size and be resorbable with appropriate biodegradation rates.Poly(a-caprolactone) (PCL) is a kind of materials suit for the demands above. PCL is biocompatible and biodegradable aliphatic polyester which is nontoxic for living organisms and bioresorbable after a period of implantation. Because of its unique combination of biocompatibility, permeability and biodegradability, PCL and some of its copolymer with lactides and glycolide have been widely applied in medicine as artificial skin, artificial bone and containers for sustained drug release.Goel and Beckman have reported a new method to generate microcellular poly(methy l methacrylate) foams in which the samples are saturated with CO2 under a series of supercritical (SC)conditions, and then the system is rapidly depressurized to atmospheric pressure at constant temperature. Unlike traditional methods, it reduces glass-transition temperature (Tg) of the mixture to below the experimental temperature rather than directly heat the system above Tg. In this process of nucleation, no phase separation occurs as well as no phase boundary meets, so the cellular structure of the foam can be retained better.In this work, we have generated PCL foams by using supercritical CO2. Because of the low glass transition temperature (Tg = -60 ℃) of PCL far below the ice point, the experimental temperature in our study is much higher than Tg, which is different from the studies by others before. A series of variable factors on the foam structure as saturation temperature, saturation pressure, saturation time and depressurization

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

Early response of pressurized hot water in a pipe to a sudden break. Final report

International Nuclear Information System (INIS)

Alamgir, M.; Kan, C.Y.; Lienhard, J.H.

1981-06-01

Experimental and analytic studies that explain the details of early pressure variations during rapid depressurization in water-cooled reactors are presented as a means of assessing sudden break consequences in a coolant pipe. The report includes (1) a description of the experiment, (2) an analysis of the new bubble growth law for thermally controlled growth of vapor bubbles in an exponentially-varying pressure field, and (3) a review of previous studies and additional observations of blowdown behavior

Blow.MOD2: a program for blowdown transient calculations

International Nuclear Information System (INIS)

Doval, A.

1990-01-01

The BLOW.MOD2 program has been developed to calculate the blowdown phase in a pressurized vessel after a break/valve is opened. It is a one volume model where break height and flow area are specified. Moody critical flow model was adopted under saturation conditions for flow calculation through the break. Heat transfer from structures and internals have been taken into account. Long term depressurization results and a more complex model are compared satisfactorily. (Author)

Generic evaluation of small break loss-of-coolant accident behavior in Babcock and Wilcox designed 177-FA operating plants

International Nuclear Information System (INIS)

1980-01-01

Slow system depressurization resulting from small break loss-of-coolant accidents (LOCAs) in the reactor coolant system have not, until recently, received detailed analytical study comparable to that devoted to large breaks. Following the TMI-2 accident, the staff had a series of meetings with Babcock and Wilcox (B and W) and the B and W licensees. The staff requested that B and W and the licensees: (1) systematically evaluate plant response for small break loss-of-coolant accidents; (2) address each of the concerns documented in the Michelson report; (3) validate the computer codes used against the TMI-2 accident; (4) extend the break spectrum analysis to very small breaks, giving special consideration to failure of pressurizer valves to close; (5) analyze degraded conditions where AFW is not available; (6) prepare design changes aimed at reducing the probability of loss-of-coolant accidents produced by the failure of a PORV to close; and (7) develop revised emergency procedures for small breaks. This report describes the review of the generic analyses performed by B and W based on the requests stated above

Method for operating nuclear reactor

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

1978-01-01

Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)