APR1400 CEA Withdrawal at Power Accident Analysis using KNAP

International Nuclear Information System (INIS)

Lee, Dong-Hyuk; Yang, Chang-Keun; Kim, Yo-Han; Sung, Chang-Kyung

2006-01-01

KEPRI (Korea Electric Power Research Institute) has been developing safety analysis methodology for non- LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code. RETRAN code is a non- LOCA safety analysis code developed by EPRI. The new methodology will replace existing CE (Combustion Engineering) supplied codes and methodologies currently used in non-LOCA analysis of OPR1000. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400). The CEA (Control Element Assembly) withdrawal at power accident is one of the 'reactivity and power distribution anomalies' events and the results are typically described in the chapter 15.4.2 of SAR (Safety Analysis Report). The APR1400 has been designed to generate 1,400MWe of electricity with advanced features for greatly enhanced safety and economic goals. The CEA withdrawal at power analysis in APR1400 SSAR (Standard Safety Analysis Report) is analyzed with CESEC-III computer code. In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, CEA withdrawal at power accident is analyzed using RETRAN code and it is compared with results from APR1400 SSAR

Design characteristics of EU-APR1400 on-site power system

International Nuclear Information System (INIS)

Kim, D.H.; Kim, Y.S.; Kim, Y.S.

2014-01-01

In the global nuclear market, US and European design requirements have been largely used to develop the design of nuclear power plants(NPPs). The APR1400 design was developed on the basis of US regulatory guide and EPRI utility requirements document(URD). In order to enlarge the export market of APR1400, KHNP (Korea Hydro & Nuclear Power Co., Ltd) has developed the EU-APR1400 design which complies with the European nuclear design requirements. In this paper, the design characteristics of EU-APR1400 on-site power system developed according to the European design requirements of electrical power system are described. The European main design requirements of electrical power system involve 50 Hz rated frequency, 400/110 kV grid voltage, the application of the diversity and the redundancy, and so on. The EU-APR1400 on-site power system has been developed on the basis of these requirements. The representative designs include the redundancy, diversity, independence design, the emergency power supply design, the design for providing electrical power to the dedicated severe accident systems, and the design for European grid requirements. (author)

System Response Analysis of LOCV for APR1400 Using KNAP

International Nuclear Information System (INIS)

Kim, Yo-Han; Sung, Chang-Kyung; Jun, Hwang-Yong

2006-01-01

Korea Electric Power Research Institute (KEPRI) of Korea Electric Power Corporation (KEPCO) has been developed the non-loss-of-coolant accident (non-LOCA) analysis methodology, called as the Korea Non-LOCA Analysis Package (KNAP), for the typical Optimized Power Reactor 1000 (OPR1000) plants. Considering current licensing methodology conducted by ABB-CE, however, the KNAP could be applied to Advanced Power Reactor 1400 (APR1400). In spite of some difference in components of two plant types, there is a close resemblance between their nuclear steam supply systems (NSSS). So, in this study, the loss of condenser vacuum (LOCV) event were analyzed using KNAP for APR1400 to estimate the feasibility of the application and the results were compared with those given in APR1400 Standard Safety Analysis Report (SSAR), which were calculated using the CESEC-III code of ABB-CE. Through the study, it was concluded that the KNAP could be applicable to APR1400 on the view point of LOCV event

Functional Requirement Analysis and Function Allocation for APR 1400

International Nuclear Information System (INIS)

Nuraslinda, Anuar; Florah, Kamanja; Noloyiso, Mtoko and others

2013-01-01

This paper intends to fulfill the FRA and FA of the HFE as required in Chapter 4 of NUREG-0711 rev. 3 for APR1400 to satisfy both plant safety and power generation objectives. This paper aims to evaluate the FRA and FA for APR1400. The allocation of function is done at the system level for all processes for both the power generation and safety goals, following the NUREG/CR-3331 guideline. As a conclusion, this paper has successfully implemented the requirements and methodology specified in NUREG-0711 for APR 1400. The Functional Requirement Analysis (FRA) and Function Allocation (FA) are required by the regulation in the Human Factors Engineering (HFE) program. The FRA defines the functions, processes, and system for plant safety and power generation. The FA allocates the functions to human operator, automation, or a combination of two. The FRA and FA for APR1400 have been performed in the very early stage of development but only for the plant safety. However, the analysis did not include the goal of power generation and also did not fully satisfy the latest revision of NUREG-0711

APR1400 Fluidic Device Sensitivity Test

International Nuclear Information System (INIS)

Choi, Nam Hyun; Chu, In Cheol; Min, Kyong Ho; Song, Chul Hwa

2005-12-01

In the safety injection tank at the emergency core cooling system of APR1400, a new safety design feature, passive fluidic device is equipped which includes no active driving system. It is essential to evaluate the new design feature with various experiments. For this reason, three categories of sensitivity tests have been performed in the present study. As the first sensitivity experiment, the effect of the height of the stand pipe was investigated. The second sensitivity test was conducted with removing the insert plate gasket to examine its effect. The effect of the expansion of the control nozzle width was ascertained from the third sensitivity test. The results of each test showed that the passive fluidic device which will be equipped in the SIT tank of APR1400 has great integrity and repeatability

A Probability Analysis of the Generating Cost for EU-APR1400 Single Unit

International Nuclear Information System (INIS)

Ha, Gak Hyeon; Kim, Sung Hwan

2014-01-01

The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, reducing CO 2 emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the EU-APR100 development is to ensure export competitiveness in the European countries. To this end, EU-APR1400 have been developed based one te APR1400 (Advanced Power Reactor, GEN Type) The EU-APR1400 adds many advanced design features to its predecessor, as outlined below in Table 1. In this simulation, the results of the generating cost of the EU-APR1400 single unit were determined using the probability cost analysis technique, the generating cost range was shown to be 56.16 ∼ 70.92 won/kWh.

A Probability Analysis of the Generating Cost for EU-APR1400 Single Unit

Energy Technology Data Exchange (ETDEWEB)

Ha, Gak Hyeon; Kim, Sung Hwan [KHNP CRI, Seoul (Korea, Republic of)

2014-10-15

The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, reducing CO{sub 2} emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the EU-APR100 development is to ensure export competitiveness in the European countries. To this end, EU-APR1400 have been developed based one te APR1400 (Advanced Power Reactor, GEN Type) The EU-APR1400 adds many advanced design features to its predecessor, as outlined below in Table 1. In this simulation, the results of the generating cost of the EU-APR1400 single unit were determined using the probability cost analysis technique, the generating cost range was shown to be 56.16 ∼ 70.92 won/kWh.

Current status of integrated HFE V and V system of APR1400

International Nuclear Information System (INIS)

Lee, M. S.; Hong, J. H.; Lee, S. H.; Lee, Y. K.

2006-01-01

The first Advanced Light Water Reactor called Advanced Power Reactor (APR1400) has many specific features such as passive safety features, digital I and C, and digitalized main control room design. This digital-based neo-concept main control room design needs comprehensive verification and validation process to get the license for construction. An integrated system for APR1400 MMIS design validation is under development by KEPRI with the participating of KHNP. The facility which is based on the NUREG/CR-6393 and ANS3.5 can simulate all of the dynamic functions and responses of APR1400, and measure the human factors by using conventional subject methods and measuring the operator's biological signals. This system will be used in human factors V and V process of the PRM (NUREG0711) of APR1400 MMIS design. (authors)

Study on Maintenance Personnel Development Plan For The Exported APR1400 Commissioning

International Nuclear Information System (INIS)

Cho, Sungbae; Kim, Jongdae; Jun, Hokwang; Hwang, Inok; Kang, Jaeyuel

2012-01-01

This paper indicates ways to develop maintenance personnel for the exported APR1400 commissioning. The exported APR1400 has not been experienced ay maintenance yet, and requirements for maintenance personnel have not been clarified yet. Based on sound maintenance experience, KEPCO Plant Service and Engineering Company (KEPCO KPS) has studied on maintenance training and career requirement to establish a development plan of the maintenance personnel for the exported nuclear power plant. By defining manpower and training requirement, and mobilization plan, we expect to secure reliability of the exported APR1400

Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

International Nuclear Information System (INIS)

Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

2013-01-01

A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

GOTHIC Simulation of APR1400 Auxiliary Charging Pump room heat up

Energy Technology Data Exchange (ETDEWEB)

Ha, Hui-Un; Heo, Sun [KHNP, Daejeon (Korea, Republic of)

2014-10-15

As a part of the Advanced Power Reactor 1400 (APR1400) U. S. Nuclear Regulatory Commission Design Certification (NRC DC) project, we have been investigating Auxiliary Charging Pump (ACP) room heat up. With reference to the design specification of the ACP room, we determined input information and developed a GOTHIC model of the APR1400 ACP room. This calculation model is described herein, and representative results from the calculation are presented as well. The results of the present paper are used to determine the integrity of ACP operating in the accident. APR 1400 GOTHIC model was developed for ACP room heat up calculation. Calculation results confirm that door opening is cooling the room properly. It is found that the difference due to the surface option of heat conductors is insignificant. Based on this result, further studies should be performed to confirm integrity of ACP.

Design features of Advanced Power Reactor (APR) 1400 steam generator

International Nuclear Information System (INIS)

Park, Tae-Jung; Park, Jun-Soo; Kim, Moo-Yong

2004-01-01

Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80 + , which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80 + steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80 + . Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

Implementation and Analysis for APR1400 Soft Control System

International Nuclear Information System (INIS)

2015-01-01

Due to the rapid advancement of digital technology, the definite technical advantages of digital control system compared to analog control system are accelerating the implementation of advanced distributed digital control system in the nuclear power plant. One of the major advantages of digital control system is the capability of Soft Control System. The design of Soft Control System for Advanced Power Reactor 1400 (APR1400) plant of Man-Machine Interface System (MMIS) is based on full digital technologies to enhance reliability, operability and maintainability. Computer-based compact workstation has been adopted in the APR1400 Main Control Room (MCR) to provide convenient working environment. This paper introduces the approaches and methodologies of Soft Control System for the Advanced Control Room (ACR). This paper also explains major design features for operation and display of the Soft Control System and its implementation to cope with regulatory requirements. (authors)

Conceptual Designs for the Performance Improvement of APR1400 SIT and Preliminary Performance Evaluation

International Nuclear Information System (INIS)

Chu, In-Cheol; Kwon, Tae-Soon; Song, Chul-Hwa

2008-01-01

Some evolutionary type PWRs such as APR1400 and APWR adopt advanced safety injection tank (SIT). The SIT of APR1400 has a fluidic device (FD) which passively controls ECC water injection flow rate into reactor coolant system during refill and reflood phases of LB-LOCA (i.e., a high injection flow rate during the refill phase and a low injection flow rate during the reflood phase). The benefit of the FD is the elimination of the function of low pressure safety injection pump from the safety injection system. The flow controlling performance of the APR1400 FD was evaluated using a prototypical full-scale test facility, called VAPER (Valve Performance Evaluation Rig). Even though the performance of the APR1400 FD satisfied major design and licensing requirements, further improvement of the performance is expected such as the extension of total injection period, the delay of nitrogen gas discharge. Several conceptual designs have been being drawn out in order to improve the performance of the APR1400 SIT. The performance of some designs was evaluated using a small scale SIT test rig. The present paper introduces some of the conceptual designs and shows the performance evaluation experimental results

Non-Integrated Standalone Tests of APR1400 Simulator

International Nuclear Information System (INIS)

Hwang, Su Hyun; Lee, Jeong Ik; Hong, Soon Joon; Lee, Byung Chul; Seo, Jeong Gwan; Lee, Myung Soo

2007-01-01

APR1400 being developed for the construction of New Kori 3 and 4 Units has improved safety and more economical efficiency compared with previous PWR. The ESF(Engineered Safety Features) newly introduced to enhance safety are as follows: DVI (Direct Vessel Injection), Fluidic Device, IRWST (In-containment Refueling Water Storage Tank). So the transient pattern of anticipated accidents will show different characteristics from previous PWR. There are multidimensional flow phenomena like as emergency core cooling coolant bypass discharge in the downcomer, downcomer boiling, and different safety injection characteristics due to fluidic device during LBLOCA. Also there is the phenomenon of critical flow due to the open of pressurizer POSRV (Pilot Operated Safety Relief valve) connected to IRWST and safety depressurization system and the prediction of discharge flow is very important. KEPRI is developing APR1400 simulator using RELAP-RT . RELAP-RT was developed by DS and S (Data systems and Solutions) based on RELAP5/MOD3.2. The improved features of RELAP-RT to function as a simulator are as follows: Add simulator functionality - Control by simulator executive - IC snap and reset capability - Back-track snap and reset capability - Fast time capability. Fast time capability(examples) - The rate of condensation has been limited. - Fictional choking model has been developed for internal junctions. - Wall heat transfer coefficients and heat fluxes has been limited. In this study, various NISTs (Non-Integrated Standalone Tests) were performed to verify the capability of RELAP-RT as APR1400 simulator by the comparison with RELAP5/MOD3.3

System Response Analysis of Rod Ejection Accident for APR1400 Using KNAP Hot Spot Model

International Nuclear Information System (INIS)

Kim, Yo-Han; Ha, Sang-Jun; Jun, Hwang-Yong

2006-01-01

Korea Electric Power Research Institute (KEPRI) has been developed the non-loss-of-coolant accident (non- LOCA) analysis methodology, called as the Korea Non- LOCA Analysis Package (KNAP), for the typical Optimized Power Reactor 1000 (OPR1000) plants. Considering current licensing methodology conducted by ABB-CE, however, the KNAP could be applied to Advanced Power Reactor 1400 (APR1400) also. In spite of some difference in design concepts of two plant types, there is a close resemblance between their nuclear steam supply systems (NSSS). So, in this study, the rod ejection accident (REA) event was analyzed using KNAP hot spot model (HSM) for APR1400 to estimate the feasibility of the application and the results were compared with those given in APR1400 Standard Safety Analysis Report (SSAR), which were calculated using the CESEC-III and STRIKIN-II code of ABB-CE. Through the study, it was concluded that the KNAP could be applicable to APR1400 on the view point of REA

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

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.

Preliminary Results Of LOCA Problem For APR1400 Reactor

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Le Thi Thu; Vo Thi Huong; Le Van Hong

2011-01-01

Several features of NPP with APR1400 nuclear reactor during a loss of coolant accident (LOCA) are investigated in this study. The report describes some main design characteristics of an engineering safety systems of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Large Break LOCA accident has been analyzed and evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with broken area of 0.0465 m 2 in case of high pressure safety injection system (HPSI) failed to operate or 2 and 4 HPSI pumps are activated. The preliminary results of this work is a part of collaboration between INST researchers and KAERI experts in using RELAP tool for safety analysis of NPPs. (author)

A Source Term Calculation for the APR1400 NSSS Auxiliary System Components Using the Modified SHIELD Code

International Nuclear Information System (INIS)

Park, Hong Sik; Kim, Min; Park, Seong Chan; Seo, Jong Tae; Kim, Eun Kee

2005-01-01

The SHIELD code has been used to calculate the source terms of NSSS Auxiliary System (comprising CVCS, SIS, and SCS) components of the OPR1000. Because the code had been developed based upon the SYSTEM80 design and the APR1400 NSSS Auxiliary System design is considerably changed from that of SYSTEM80 or OPR1000, the SHIELD code cannot be used directly for APR1400 radiation design. Thus the hand-calculation is needed for the portion of design changes using the results of the SHIELD code calculation. In this study, the SHIELD code is modified to incorporate the APR1400 design changes and the source term calculation is performed for the APR1400 NSSS Auxiliary System components

Extended station blackout analyses of an APR1400 with MARS-KS

Directory of Open Access Journals (Sweden)

Kim Woongbae

2016-01-01

Full Text Available The Fukushima Daiichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electric energy required for essential systems during a station blackout is provided from 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 extended station blackout scenarios, the best estimate MARS-KS computer code 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 on reactor coolant pump seal leakage was carried out.

Verbal Communication in the APR1400 Advanced Control Room

International Nuclear Information System (INIS)

Kim, Jong Hyun; Shin, Young Cheol

2007-01-01

This paper introduces the characteristics of communication in advanced main controlghd rooms (MCRs) and some observations from a case study performed for APR1400 MCR. In advanced MCRs, operators need not maintain the same communication patterns as they do in conventional ones. For example, a senior reactor operator (SRO) does not have to rely on board operators for information acquisition and can get any information from his/her own workstation. This situation may also bring about new problems in MCRs such as lack of shared situation awareness and collaboration between MCR operators. To cope with these problems, the APR1400 MCR adapts several approaches in design and training for encouraging operators to communicate with each other. This paper introduces the possible changes of communication patterns and the countermeasures in design and training. Some findings from an integrated system validation for Shin Kori Units 3 and 4 are also presented

Seismic analysis of APR1400 RCS for site envelope using big mass method

International Nuclear Information System (INIS)

Kim, J. Y.; Jeon, J. H.; Lee, D. H.; Park, S. H.

2002-01-01

One of design concepts of APR1400 is the site envelope considering various soil sites as well as rock site. The KSNP's are constructed on the rock site where only the translational excitations are directly transferred to the plant. On the other hand, the rotational motions affect the responses of the structures in the soil cases. In this study, a Big Mass Method is used to consider rotational motions as excitations at the foundation in addition to translational ones to obtain seismic responses of the APR1400 RCS main components. The seismic analyses for the APR1400 excited simultaneously by translation and rotational motions were performed. The results show that the effect of soil sites is not significant for the design of main components and supports of the RCS, but it may be considerable for the design of reactor vessel internals, piping, and nozzles which have lower natural frequencies

Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

International Nuclear Information System (INIS)

Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

2007-01-01

APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

Containment Performance Analysis with Large Break LOCA for EU-APR1400

International Nuclear Information System (INIS)

Hwang, Do Hyun; Lee, Keun Sung; Kim, Yong Soo

2013-01-01

In this paper for containment performance analysis, the containment pressurization analysis is performed and thermo-hydraulic response analysis of containment structure is carried out to provide basic understanding of containment transient states under a severe accident sequence. Especially, in EU-APR1400 design, to reduce containment pressure and temperature, Severe Accident Containment Spray System (SACSS) is designed to be actuated automatically when Core Exit Temperature (CET) reaches 922 K (649 .deg. C). The containment performance analysis was carried on LBLOCA sequence for EU-APR1400 with SACSS through MAAP code. If SACSS is actuated when CET reaches 922 K (649 .deg. C) , the containment pressure and temperature decrease to a sufficient low level. The predicted atmospheric pressure of containment will not exceed the ultimate pressure capacity (UPC) and have a sufficient margin to it even though the UPC of the reference plant (Shin-Kori Units 3 and 4) is used instead because the UPC calculation for EU-APR1400 has not been completed. The largest load on the containment by LBLOCA is estimated at 306.1 kPa. Thus the margin to UPC is estimated to be 330 % in comparison with 1.329 MPa as UPC for the reference plant.

Features of Computerized Procedure System of Shin-Kori unit 5 and 6

International Nuclear Information System (INIS)

Seong, Nokyu; Jung, Yeonsub; Sung, Chanho

2016-01-01

The Computerized Procedure System (CPS) is one of the Man Machine Interface (MMI) resources of Main Control Room (MCR) of the Advanced Power Reactor 1400 (APR1400). The CPS has been continuously improved since it was installed in Shin-Kori unit 3 and 4. The Korea Hydro Nuclear Power Central Research Institute (KHNP CRI) has found the points of improvement of CPS through CPS centered Human Factors Engineering Verification and Validation (HFE V and V) and Operating Experience Review (OER) of reference power plant. This paper shows the main features of CPS of Shin-Kori 5 and 6 unit. This paper shows the main features of CPS of Shin-Kori 5 and 6. These are some of improvements of CPS. This prototype of CPS currently is implementing in CRI. The respective function can be more detailed after testing the prototype. These features will be applied to Shin-Kori 5 and 6 CPS after HFE V and V

Features of Computerized Procedure System of Shin-Kori unit 5 and 6

Energy Technology Data Exchange (ETDEWEB)

Seong, Nokyu; Jung, Yeonsub; Sung, Chanho [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Computerized Procedure System (CPS) is one of the Man Machine Interface (MMI) resources of Main Control Room (MCR) of the Advanced Power Reactor 1400 (APR1400). The CPS has been continuously improved since it was installed in Shin-Kori unit 3 and 4. The Korea Hydro Nuclear Power Central Research Institute (KHNP CRI) has found the points of improvement of CPS through CPS centered Human Factors Engineering Verification and Validation (HFE V and V) and Operating Experience Review (OER) of reference power plant. This paper shows the main features of CPS of Shin-Kori 5 and 6 unit. This paper shows the main features of CPS of Shin-Kori 5 and 6. These are some of improvements of CPS. This prototype of CPS currently is implementing in CRI. The respective function can be more detailed after testing the prototype. These features will be applied to Shin-Kori 5 and 6 CPS after HFE V and V.

Effectiveness of External Reactor Vessel Cooling (ERVC) strategy for APR1400 and issues of phenomenological uncertainties

International Nuclear Information System (INIS)

Oh, S.J.; Kim, H.T.

2007-01-01

The APR1400(Advanced Power Reactor 1400) is an evolutionary advanced light water reactor with rated thermal power of 4000 MWt. For APR1400, External Reactor Vessel Cooling (ERVC) is adopted as a primary severe accident management strategy for in-vessel retention (IVR) of corium. The ERVC is a method of IVR by submerging the reactor vessel exterior. At the early stage of the APR1400 design, only ex-vessel cooling, cooling of the core melt outside the vessel after vessel is breached, is considered based on the EPRI Utility Requirement Document for Evolutionary LWR. However, based on the progress in implementation of Severe Accident Management Guidance (SAMG) for operating plants, as well as the research findings related to ERVC, ERVC strategy is adopted as a part of key severe accident management strategies. To improve its success, the strategy is reviewed and we implemented necessary design arrangement to increase its usefulness in managing the severe accident. In this paper, we examine the evolution of ERVC concept and its implementation in APR1400. Then, we review possible approach, including Risk-Oriented Accident Analysis Methodology (ROAAM), to evaluate the effectiveness of the strategy. (authors)

MDEP Design-Specific Common Position CP-APR1400WG-01. Common position addressing Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

2016-05-01

The MDEP APR1400 Working Group (APR1400WG) members consist of members from Republic of Korea, United Arab Emirates, and the United States. A main objectives of MDEP is to encourage convergence of code, standard and safety goals with exploring the opportunities for harmonization of regulatory practice and cooperation on safety review of APR-1400 specific designs. This common position addressing is aimed at sharing knowledge, information and experience on safety improvement related to lessons learned from the Fukushima Daiichi NPP Accident or Fukushima Daiichi NPP Accident-related issues amongst APR-1400 WG member states to achieve the MEDP goal. Because not all of these Regulators have completed the regulatory review of their APR1400 applications yet, this paper identifies common preliminary approaches to address potential safety improvements for APR1400 plants, as well as common general expectations for new nuclear power plants, as related to lessons learned from the Fukushima Daiichi NPP Accident or Fukushima Daiichi NPP Accident-related issues. While some asymmetry exists among those of three Regulators in terms of design, regulatory practice and licensing milestone sharing information and common understanding on post-Fukushima Daiichi NPP Accident enhancement would be promote resilient design for countering beyond design extreme external event like Fukushima Daiichi NPP nuclear disaster. This common position paper aims at identifying characteristics of post-Fukushima Daiichi NPP Accident enhancements putting in place by each country and setting common position to achieve balanced and harmonized APR-1400 design. After the safety reviews of the APR1400 design applications that are currently in review are completed, the regulators will update this paper to reflect their safety conclusions regarding the APR1400 design and how the design could be enhanced to address Fukushima Daiichi NPP Accident-related issues. The common preliminary approaches are organised into

Analysis of a 12-Finger Rod Drop using RETRAN/MASTER Code System for APR1400

International Nuclear Information System (INIS)

Yu, Keuk Jong; You, Sung Chang; Kim, Han Gon

2009-01-01

The Optimized Power Reactor 1000 (OPR1000) has 4-finger and 12-finger Control Element Assemblies (CEAs). When the 12-finger CEA is dropped, Core Protection Calculator System (CPCS) shuts down the reactor to prevent fuel damage that could occur from the sudden reactor power peaking. By contrast, the improved CPCS of Advanced Power Reactor 1400 (APR1400), which has systems similar to those of the OPR1000, decreases reactor power rapidly using its Reactor Power Cutback System (RPCS) to avoid unwanted reactor trips caused by the CPCS during a 12- finger CEA drop event. RETRAN is a best-estimate code for transient analysis of Non-LOCA. The RETRAN control logic, which includes the function of reducing reactor power during a 12-Finger CEA drop, has been developed for the APR1400. A MATRAN program has also been developed. MATRAN is the interface program for realtime processing to connect RETRAN with MASTER code which is a nuclear analysis and design code. MATRAN supplies adequate feedback reactivities from the MASTER code to RETRAN code. The purpose of this study is to analyze the behavior of a nuclear reactor core and its primary system using conventional RETRAN analysis procedure and MATRAN program analysis procedure during a 12- finger CEA drop. In addition, the axial power distribution and Axial Shape Index (ASI) are produced by the MATRAN program and they are confirmed as within operation limits

Application of Nuclear Application Programs to APR1400 Simulator

International Nuclear Information System (INIS)

Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk

2012-01-01

Advanced Power Reactor 1400MWe (APR1400) simulator has been developed and installed at Kori Training Center for operators of ShinKori no.3, 4 nuclear power plant by Korea Hydro and Nuclear Power,s Central Research Institute (KHNP CRI). NAPS (Nuclear Application Programs) is a computerbased system which provides operators with past and real-time information for monitoring and controlling NSSS (Nuclear Steam Supply System), BOP (Balance Of Plant) and Electric system. NAPS consists of several programs such as COLSS (Core Operating Limit Supervisory System), SPADES+ (Safety Parameter Display and Evaluation System), CEA (Control Element Assembly) Application Program, and so on. Each program makes calculations based on its own algorithm and provides information available for operation. In order to use NAPS programs with a simulator even though they are being used in a real plant, they should be modified to add several simulation functions such as reset, snap, run/freeze and backtrack required by ANSI/ANS-3.5 to the original NAPS functionality. On top of that, interfacing programs should be developed for the data communication between respective NAPS programs and simulator sever. The purpose of this paper is to provide the overall architecture of the communication system between NAPS and simulator model, and to describe the method to apply NAPS to APR1400 simulator

Critical heat flux for APR1400 lower head vessel during a severe accident

International Nuclear Information System (INIS)

Noh, Sang W.; Suh, Kune Y.

2013-01-01

Highlights: ► Studied boiling on downward-facing hemispherical vessel with asymmetric thermal insulator. ► Scaled the APR1400 lower head linearly down by 1/10 including ICI tubes and shear keys. ► Performed thermal analysis using ANSYS V11.0 to determine the internal temperature and heat flux. ► Performed tests to obtain the CHF with saturated demineralized water at atmospheric pressure. ► Measured CHF accounting for 3D random flow effect expected in the APR1400 application. -- Abstract: Corium Ablation Stopper Apparatus (CASA) has a downward-facing hemispherical vessel and geometrically asymmetric thermal insulator of the Advanced Power Reactor 1400 MWe (APR1400) scaled linearly down by 1/10, as well as sixty-one in-core instrumentation (ICI) tubes and four shear keys. The heated vessel plays a pivotal role in CASA depending on the configuration of the oxide pool and metal layer to bring about the focusing effect expected of a molten pool in the lower head during a severe accident. The heated vessel was designed through a trial-and-error method and thermal analysis. Thermal analysis was performed using ANSYS V11.0 to investigate the effect of the internal temperature and heat flux on the integral hemispherical copper vessel. The CASA tests were carried out to obtain the critical heat flux (CHF) with saturated and demineralized water at the atmospheric pressure (0.1 MPa). The CHF in the metal layer through the hemispherical channel was found to be lower than that in the ULPU-2400 configuration V data through the streamlined thermal insulator. The experimental CHF was measured and obtained through the CASA hemispherical heated surface accounting for the three-dimensional random flow effect expected in the APR1400 application

Quality Control of Concrete Structure For APR1400 Construction

International Nuclear Information System (INIS)

Seo, Inseop; Song, Changhak; Kim, Duill

2012-01-01

Nuclear structure shall be constructed to protect internal facilities in the normal operation against external accidents such as the radiation shielding, earthquakes and to be leak-proof of radioactive substances to the external environment in case of loss of coolants. containment and auxiliary building of nuclear power plants are built in reinforced concrete structures to maintain these protection functions. Nuclear structures shall be designed to ensure soundness in operation since they are located on the waterfront where is easy do drain the cooling water and so deterioration and damage of concrete structures caused by seawater can occur. Durability is ensured for concrete structures of APR1400, a Korea standard NPP, in compliance with all safety requirements. In particular, owners perform quality control directly on the production and pouring of cast in place concrete for the concrete structure construction to make sure concrete structures established with quality homogeneity and durability. This report is to look into the quality control standard and management status of cast in place concrete for APR1400 construction

APR1400 Electrical Power System Conformance to SECY-91-078

Energy Technology Data Exchange (ETDEWEB)

Ha, Che-Wung; Kim, Yun-Ho [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In this paper, the APR1400 electric power system for NRC-DC was described. In addition, the conformance to SECY 91-078 was evaluated. Enclosure 1 of the SECY provides an overview of the issue and states that the staff concludes that feeding the safety buses from the offsite power sources through non-safety buses or from a common transformer winding with non-safety loads is not the most reliable configuration. Such an arrangement increases the difficulty in properly regulating voltage at the safety buses, subjects the safety loads to transients caused by the non-safety loads, and adds additional failure points between the offsite power sources and safety loads. Therefore, it is the staff’s position that at least one offsite circuit to each redundant safety division should be supplied directly from one of the offsite power sources with no intervening non-safety buses, in such a manner that the offsite source can power the safety buses upon a failure of any non-safety bus. The APR1400 does not have an intervening nonsafety bus in the current offsite to onsite electrical configuration; however, the design does include nonsafety and safety buses coming from the same secondary side 4.16 kV transformer winding. Nevertheless, the APR1400 has designed the electrical interface system between offsite and onsite power with enhanced reliability measures to ensure that the nonsafety system will not impact the safety loads. The design complies with GDC 17 and also conforms to SECY-91-078.

Analytical Assessment of Environmental Impact for APR1400DC UHS Cooling Tower

Energy Technology Data Exchange (ETDEWEB)

Lee, Jaiho [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Hot process water is pumped from the plant process to the cooling towers. Heat is rejected through evaporation of the process water, interacting with ambient air blown upward by fans.. Plumes generated from exit ports of the cooling tower may have adverse effects on the environment, such as deposition of cooling tower drift release, fogging, icing, shadowing, and ground-level temperature and humidity increase. These kinds of environmental impact of the cooling tower are linked closely with the dispersion of the cooling tower plumes. In this respect, predicting the behavior of the plumes has become one of the most important issues in the environmental assessments of the cooling towers. The SACTI (seasonal/annual cooling tower impact) model is an analytical tool to predict the environmental effect of cooling tower, which was developed by Argonne National Laboratory and University of Illinois with support from EPRI (electric power research institute). The initial version of SACTI has been widely used to assess the environmental effect of cooling towers in many industrial fields such as steam power plants and NPPs. Guo et. al. investigated impact of heat rejection and cooling tower height on plume dispersion using the SACTI model, for the purpose of the future construction of inland NPPs. They found that increasing cooling tower height decreases the plume length and height frequencies. Their simulation results showed that the increase in heat rejection increases the plum radius frequency. The APR1400DC is an advanced light water reactor developed for the purpose of NRC-DC (design certification). The cooling towers for APR1400DC UHS consist of two linear mechanical draft cooling towers (LMDCTs). The LMDCT for APR1400DC UHS is conceptually designed because the plant site has not been decided yet. In the present study, the dependency of plume dispersion on the number of cooling towers is investigated using SACTI-2-beta, for predicting annual environmental effect of APR

Structural assessments of plate type support system for APR1400 reactor

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Anh Tung; Namgung, Ihn, E-mail: inamgung@kings.ac.kr

2017-04-01

Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

Numerical Analysis of Molten Corium Dispersion during Hypothetical High-Pressure Accidents in APR1400 Nuclear Power Plant

International Nuclear Information System (INIS)

Kim, Jong Tae; Ha, Kwang Soon; Kim, Sang Baik; Kim, Hee Dong; Jeong, Jae Sik

2010-01-01

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by the following jet of a high pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet with very high velocity and is released into the upper compartment of the NPP by an overpressure in the cavity. The heat-carrying fragments of the corium transfer the thermal energy to the ambient air in the containment and react chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. New generation NPPs such as APR1400 and EPR have been designed in consideration of reducing the possibility of the containment failure from the DCH. In order for that, APR1400 has a convolute-type corium chamber connected to the reactor cavity. In the case of EPR, severe-accident dedicated depressurization valves are installed to preclude a high pressure melt ejection (HPME). DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical reaction. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. The corium dispersion rates for many types of the NPP containments had been obtained by experiments in 90s. And some correlations from the experimental data were developed. As mentioned above, APR1400 has a corium chamber to reduce the corium dispersion rate. But there is no experimental data for the dispersion rate specific to the APR1400 cavity geometry. So its performance for capturing of the dispersed corium

An Experimental Investigation on APR1400 Penetration Weld Failure by Metallic Melt

International Nuclear Information System (INIS)

An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol

2014-01-01

The penetrations are considered as the most vulnerable parts with respect to the reactor vessel failure when a core melt severe accident occurs and the corium reaches the lower head. Penetration tube failure modes can be divided into two categories; tube ejection out of the vessel lower head and rupture of the penetration tube outside the vessel. Tube ejection begins with degrading the penetration tube weld strength to zero as the weld is exposed to temperatures as high as the weld melting temperature, which is called weld failure, and then overcoming any binding force in the hole in the vessel wall that results from differential thermal expansion of the tube and vessel wall. Tube rupture assumes that the debris bed has melted the instrument tube inside the reactor and melt migrates down into the tube to a location outside the vessel wall where a pressure rupture can occur, thus breaching the pressure boundary. In the present paper, we have a focus on the tube ejection failure mode, specifically on the APR1400 weld failure by direct contact with a metallic melt. The objective is to investigate experimentally the ablation kinetics of an APR1400 penetration weld during the interactions with a metallic melt and to suggest the modification of the existing weld failure model. This paper involves the interaction experiments of two different metallic melts (metallic corium and stainless steel melts) with a weld specimen, and rough estimation of weld failure time. The interaction experiments between the metallic melts and an APR1400 penetration weld were performed to investigate the ablation kinetics of the penetration weld. Metallic corium and stainless steel melts were generated using an induction heating technique and interacted with a penetration weld specimen. The ablation rate of the weld specimen showed a range from 0.109 to 0..244 mm/s and thus the APR1400 penetration weld was estimated to be failed at hundreds of times after the interaction with the melt

Single-tube condensation experiment in Passive Auxiliary Feedwater System of APR1400+

Energy Technology Data Exchange (ETDEWEB)

Shin, Chang Wook; No, Hee Cheon; Yun, Bong Yo; Jeon, Byong Guk [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2012-05-15

Conventional Korean nuclear power plants, Advanced Power Reactors (APR), are characterized by an active cooling system. However, Active cooling system may not prevent significant damage without any AC power source available for its operation as vividly illustrated through the recent Fukushima incident. In the APR1400+ to be designed, an independent passive cooling system was added in order to overcome the aforementioned shortcomings. In the Passive Auxiliary Feedwater System (PAFS), gravity force and density difference between steam and water are used. The system comprises of 240 condensation tubes to efficiently remove decay heat. Before applying the PAFS to APR1400+, the system's safety and heat removal performance must be verified. The present study experimentally evaluates the heat removal performance of a single tube in the PAFS. The objectives of SCOP (Single-tube Condensation experiment facility of PAFS) are the evaluation of the heat removal performance in the tube of the PAFS and database construction under various tube designs and test conditions. Reaching these objectives, we developed advanced measurement techniques for the amount of moisture, heat flux, and water film thickness.

Vibration Analysis for Steam Dryer of APR1400 Steam Generator

Energy Technology Data Exchange (ETDEWEB)

Han, Sung-heum; Ko, Doyoung [KHNP CRI, Daejeon (Korea, Republic of); Cho, Minki [Doosan Heavy Industry, Changwon (Korea, Republic of)

2016-10-15

This paper is related to comprehensive vibration assessment program for APR1400 steam generator internals. According to U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 (Rev.3, March 2007), we conducted vibration analysis for a steam dryer as the second steam separator of steam generator internals. The vibration analysis was performed at the 100 % power operating condition as the normal operation condition. The random hydraulic loads were calculated by the computational fluid dynamics and the structural responses were predicted by power spectral density analysis for the probabilistic method. In order to meet the recently revised U.S. NRC RG 1.20 Rev.3, the CVAP against the potential adverse flow effects in APR1400 SG internals should be performed. This study conducted the vibration response analysis for the SG steam dryer as the second moisture separator at the 100% power condition, and evaluated the structural integrity. The predicted alternating stress intensities were evaluated to have more than 17.78 times fatigue margin compared to the endurance limit.

Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code

International Nuclear Information System (INIS)

Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu

2014-01-01

In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

Verification of SAMG entry condition for APR1400

Energy Technology Data Exchange (ETDEWEB)

Yun, Junghyun; Kim, Taewan; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-10-15

In the wake of the Fukushima accident, severe accident management has become important more than ever. While Emergency Operating Procedures (EOPs) focuses on cooling down the core during a design basis accident, the prime objective of severe accident management guideline (SAMG) is to prevent the release of radioactive material into the environment during a severe accident. Only one fixed value of core exit temperature (CET) has been applied for several decades to different types of nuclear power plants (NPPs) in Korea. Although the different types of NPPs have different cladding materials and system designs, the identical CET, i. e., 650 .deg. C, is applied as the entry condition to the SAMG of all the NPPs (except CANDU plants), i. e., Westinghouse type, OPR1000, and APR1400. In this paper, the transition point is re-evaluated for dominant severe accident sequences; station black-out (SBO), small break loss of coolant accident (SBLOCA), and me dium break loss of coolant accident (MBLOCA). In cases of SBO and SBLOCA, the current SAMG entry condition seems proper, while it needs to be reconsidered for MBLOCA case.

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

Science.gov (United States)

Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

2018-02-01

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

MELCOR code modeling for APR1400

Energy Technology Data Exchange (ETDEWEB)

Choi, Young; Park, S. Y.; Kim, D. H.; Ahn, K. I.; Song, Y. M.; Kim, S. D.; Park, J. H

2001-11-01

The severe accident phenomena of nuclear power plant have large uncertainties. For the retention of the containment integrity and improvement of nuclear reactor safety against severe accident, it is essential to understand severe accident phenomena and be able to access the accident progression accurately using computer code. Furthermore, it is important to attain a capability for developing technique and assessment tools for an advanced nuclear reactor design as well as for the severe accident prevention and mitigation. The objective of this report is to establish technical bases for an application of the MELCOR code to the Korean Next Generation Reactor (APR1400) by modeling the plant and analyzing plant steady state. This report shows the data and the input preparation for MELCOR code as well as state-state assessment results using MELCOR code.

The Gravity Makeup on the LORHR Event during Mid-loop Operation for APR1400

Energy Technology Data Exchange (ETDEWEB)

Kim, Cheol Woo; Lee, Yong Hee; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO, Daejeon (Korea, Republic of)

2015-05-15

This paper is to investigate means of the available gravity makeup after the event to prevent or sustain the core uncovery and fuel failure for the typical advanced power reactor nuclear power plants (APR1400). Unlike the OPR1000 (optimized power reactor 1000 nuclear power plants), the refueling water storage tank of APR1400 cannot be used for gravity makeup since it is located on the bottom of the containment with a lower elevation than the reactor vessel. So, the other means of gravity makeup and their effect on the core uncovery are accounted. The use of the water used for the cask loading pit (CLP) during refueling or the safety injection line filling tank (SIFT), which are not designed for gravity makeup during the event, and safety injection tanks (SITs) are considered as the alternative source of makeup feed. The results show that gravity makeup of a SIT provides a sufficient operator action time for the LORHR with the station black out (SBO) during the mid-loop operation for the APR1400. From the LORHR event analysis, the effective means of gravity makeup for APR1400 to cope with this event concurrent loss of all AC power is determined as the gravity feed using SITs. Other means for gravity makeup using CLP or SIFT, which are not designed to mitigate the LORHR event, are evaluated as not effective to the event due to the large flow resistances and the low elevation head to overcome the system pressure increase during the LORHR. The use of one SIT and two SITs for the LORHR provides the core uncovery times as 3 hr 8 min and 4 hr 32 min extending 1 hr 15 min and 2 hr 38 min from the base case, respectively.

An investigation of major influences on the seismic response of APR1400 reactor vessel internals - 15145

International Nuclear Information System (INIS)

Byun, Y.J.; Kim, J.G.; Sung, K.K.; Lee, D.H.

2015-01-01

This paper deals with 3 topics concerning the APR1400 reactor vessel internals (RVI) seismic analysis: nonlinear problems, approaches to account for uncertainties of seismic model, and dynamic responses to various seismic excitations. First, the noticeable nonlinear characteristics of the RVI seismic model are discussed, and the modeling methods for properly simulating the nonlinear behaviors of RVI under seismic loads are presented. By applying these methods to the seismic model, the seismic analysis can correctly predict the dynamic response of RVI. Next, two approaches to account for the uncertainties of seismic model are evaluated: the time history broadening method, and the sensitivity analysis based on NUREG-0800, Section 4.2, Appendix A. From the evaluation results, it is confirmed that the time history broadening method employed in the seismic analysis of APR1400 RVI sufficiently accounts for the uncertainty of seismic model. Finally, the response characteristics of APR1400 RVI to various seismic excitations are investigated. The seismic excitations corresponding to various soil profiles, including the effects of cracked and un-cracked concrete stiffness on the reactor containment building structure, are used as forcing functions. From this study, the effects of various site conditions on the dynamic response of APR1400 RVI are identified. As a result, the enveloped seismic responses obtained from this study will contribute to the development of RVI seismic design that covers a wide range of potential site conditions. (authors)

Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

Energy Technology Data Exchange (ETDEWEB)

Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

2017-03-15

Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

International Nuclear Information System (INIS)

Park, Jong-beom; Park, No-Cheol; Lee, Sang-Jeong; Park, Young-Pil; Choi, Youngin

2017-01-01

Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

International Nuclear Information System (INIS)

Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong

2013-01-01

The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft 2 ∼0.02ft 2 (DVI) 0.5ft 2 ∼0.02ft 2 (hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA

Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong [Central Research Institute, KHNP, Ltd., Daejeon (Korea, Republic of)

2013-10-15

The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft{sup 2}∼0.02ft{sup 2}(DVI) 0.5ft{sup 2}∼0.02ft{sup 2}(hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA.

Development of phenomena identification and ranking table for APR1400 main steam line break

International Nuclear Information System (INIS)

Song, J. H.; Chung, B. D.; Jeong, J. J.

2003-01-01

A Phenomena Identification and Ranking Table (PIRT) was developed for the Main Steam Line Break (MSLB) event of an APR-1400 (Advanced Power Reactor-1400). A team of experts from research institutes, industries, and regulatory bodies participated in the development. The selected event was a double-ended steam line break at full power with the reactor coolant pump running. The panel selected the fuel performance as the primary safety criterion for ranking. The plant design data, the results of APR-1400 safety analysis, and the results of additional best estimate analysis by MARS2.1 were utilized. Three phases of pre-trip, rapid cool-down, and safety injection phase were identified. Then, the ranking of a system, components, phenomenon/process based on the relative importance to the primary evaluation criterion were followed for each time phase. Finally, the knowledge-level for each important process in the component was ranked in terms of the existing knowledge. The highly ranked phenomena identified for APR-1400 MSLB are tube wall heat transfer at the steam generator shell, void distribution at the steam generator shell, liquid entrainment in the separators, mixture level in the separators, boron mixing in the upper down comer, boron transport and thermal mixing in the lower plenum, stored energy release in the upper head, and flow to and/from the upper head. The PIRT will be used as a guide in planning cost effective experimental programs and code development efforts, especially for the quantification of the process and/or phenomena, which have a high importance but low knowledge level

Application of Standardized ITAAC to the APR1400 Design Certification

Energy Technology Data Exchange (ETDEWEB)

Kang, Deogji; Kim, Yunho [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Design certification applications (DCAs) submitted pursuant to 10 CFR Part 52 contain two tiers of information, Tier 1 and Tier 2. Tier 1 contains information that is to be certified through rulemaking, and includes the inspections, tests, analyses, and acceptance criteria (ITAAC). Tier 2 contains more detailed information and is the source for information included in Tier1. DC applicants have developed the ITAAC with their own format and contents. This has caused the ineffectiveness of the review process by Nuclear Regulatory Commission (NRC) at the design certification stage. Standardizing the format and content of Tier 1 and ITAAC will achieve greater efficiency and effectiveness for Tier 1 and ITAAC development and NRC review, and for ITAAC implementation and verification. NRC provided the proposed standardized ITAAC for APR1400, and KHNP is considering the application of the proposed ITAAC. The strategy for application of the NRC proposed standardized ITAAC to the APR1400 design certification is now under developing, and a meeting with NRC to review the NRC proposed standardized ITAAC is expected to be held in the near future. In this review process, establishment of industry position on the NRC proposed standardized ITAAC is very important. The support from NEI and US industry is expected to be gained.

Application of Standardized ITAAC to the APR1400 Design Certification

International Nuclear Information System (INIS)

Kang, Deogji; Kim, Yunho

2016-01-01

Design certification applications (DCAs) submitted pursuant to 10 CFR Part 52 contain two tiers of information, Tier 1 and Tier 2. Tier 1 contains information that is to be certified through rulemaking, and includes the inspections, tests, analyses, and acceptance criteria (ITAAC). Tier 2 contains more detailed information and is the source for information included in Tier1. DC applicants have developed the ITAAC with their own format and contents. This has caused the ineffectiveness of the review process by Nuclear Regulatory Commission (NRC) at the design certification stage. Standardizing the format and content of Tier 1 and ITAAC will achieve greater efficiency and effectiveness for Tier 1 and ITAAC development and NRC review, and for ITAAC implementation and verification. NRC provided the proposed standardized ITAAC for APR1400, and KHNP is considering the application of the proposed ITAAC. The strategy for application of the NRC proposed standardized ITAAC to the APR1400 design certification is now under developing, and a meeting with NRC to review the NRC proposed standardized ITAAC is expected to be held in the near future. In this review process, establishment of industry position on the NRC proposed standardized ITAAC is very important. The support from NEI and US industry is expected to be gained

PLUS 7TM advanced fuel assembly development program for KSNPs and APR1400

International Nuclear Information System (INIS)

Kim, Kyutae; Stucker, David L.

2002-01-01

KNFC and Westinghouse have recently completed the development of the PLUS 7 TM advanced 16 X 16 fuel assembly for the Korean Standard Nuclear Plants (KSNPs) and the Advanced Power Reactor 1400 (APR 1400). This fuel design utilized the proven advanced design features including mixing vane spacer grids to increase critical heat flux performance, ZIRLO TM advanced materials to enable high-duty, high burnup fuel management and an optimized fuel rod diameter which improves fuel cycle cost while resulting in significant standardization of Korean fuel manufacture. PLUS 7 TM , also includes a patented spacer grid design with conformal fuel rod support designed to provide superior fuel rod wear/fretting resistance while minimizing pressure drop. This paper will present an overview of the PLUS 7 TM fuel assembly development process including a summary of the three-year design and testing program from a mechanical, neutronic, and thermal/hydraulic perspective. The PLUS 7 TM fuel for the KSNPs and the APR1400 reactors results in multi-million dollar per cycle savings in imported enriched uranium product for the Korean nuclear power program with technology specifically developed for Korea by experienced Korean engineers

The comparison between layout and design of APR1400 MCR and EPRI requirement

International Nuclear Information System (INIS)

Suharyo Widagdo; Darlis; Sigit Santoso; T J Suryono

2013-01-01

A good man-machine interface system is needed by a nuclear power reactor because of its complexity and high safety factor. The man-machine interface system is centralized in the main control room (MCR). All of monitoring and controlling activity to keep the installation operation safety can be done from this place. In order to increase its ability in monitoring and controlling a hard effort to raise the interface system to the highest possible level is done not only by using the latest computer technology and instrumentation system but also using Human Engineering, as well. The paper aims to compare the lay out and design the APR1400 MCR and the requirement from EPRI about main control room lay out and design as response of TMI-2 accident. We can see from the discussion that the design of the MCR of APR1400 have been applied the request of EPRI. (author)

Investigation of the Condensation Effect at IRWST Pool Surface on Containment Back Pressure in APR1400 Containment

International Nuclear Information System (INIS)

Lee, Eui Jong; Lee, Jin Yong; Lee, Byung Chul

2006-01-01

The APR1400 has several new design concepts in order to improve the plant safety functions during a postulated accident. The In-Containment Refueling Water Storage Tank (IRWST) is one of the new design concepts of APR1400 and installed at the bottom of containment building to promote the plant safety functions by simplifying emergency core cooling water source and preventing release of the fission product during an accidents. This design feature, however, brings about uncertainty factors which may necessitate conventional prediction of temperature and pressure of containment building improved or revised under accident conditions. The hot steam which is released from RCS break enters into the IRWST through four Pressure Relief Dampers (PRDs). It is expected to be condensed with water stored in IRWST, colder than incoming steam. The purpose of this study is to examine closely the influence of the condensation effect at IRWST on containment back pressure in APR1400 containment building using the GOTHIC code which can predict the steam condensation on IRWST pool surface

A Study on the Embedment Effect in the Soil-Structure Interaction Analysis of the APR1400

International Nuclear Information System (INIS)

Young-Sun Jang; Kwang-Ho Joo; Chong-Hak Kim

2002-01-01

The SSI (Soil-Structure Interaction) analyses are being performed for the APR1400 (Advanced Power Reactor 1400 MWe, Old name - KNGR; Korean Next Generation Reactor) design, because the APR1400 is developed as a Standard Nuclear Power Plant concept enveloping suitable soil conditions. For the SSI analyses, SASSI program which adopts the Flexible Volume Method is used. In the SSI analyses, there can be uncertainties by Bond and De-bond problem between the structure and lateral soil elements. According to ASCE Standard 4, one method to address this concern is to assume no connectivity between structure and lateral soil over the upper half of the embedment of 20 ft (6 m), whichever is less. This study is performed as a part of the parametric analyses for the APR1400 seismic analyses to address the concern of the potential embedment effect on the in-structure response spectra due to connectivity between structure and lateral soil. In this study, 4 model cases are analyzed to check the potential embedment effect - Full connection, 20 ft no connectivity which is defined as a minimum De-bond depth of the soil in ASCE Standard 4 and 26.5 ft no connectivity between structure and lateral soil over the upper half of the embedment. Last one is full no connection for only reference. The in-structure response spectra are compared with the response spectra without considering the embedment effect. (authors)

The effect of nodalization and temperature of reactor upper region: Sensitivity analysis for APR-1400 LBLOCA

International Nuclear Information System (INIS)

Kang, Dong Gu

2017-01-01

Highlights: • The nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature. • The effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated. • The modification of nodalization is an essential prerequisite in APR-1400 LBLOCA analysis. - Abstract: In best estimate (BE) calculation, the definition of system nodalization is important step influencing the prediction accuracy for specific thermal-hydraulic phenomena. The upper region of reactor is defined as the region of the upper guide structure (UGS) and upper dome. It has been assumed that the temperature of upper region is close to average temperature in most large break loss of coolant accident (LBLOCA) analysis cases. However, it was recently found that the temperature of upper region of APR-1400 reactor might be little lower than or similar to hot leg temperature through the review of detailed design data. In this study, the nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature, and the effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated by sensitivity analysis including best estimate plus uncertainty (BEPU) calculation. In basecase calculation, in case of modified version, the peak cladding temperature (PCT) in blowdown phase became higher and the blowdown quenching (or cooling) was significantly deteriorated as compared to original case, and as a result, the cladding temperature in reflood phase became higher and the final quenching was also delayed. In addition, thermal-hydraulic parameters were compared and analyzed to investigate the effect of change of upper region on cladding temperature. In BEPU analysis, the 95 percentile PCT used in current regulatory practice was increased due to the modification of upper region nodalization, and it occurred in the reflood phase unlike original case.

How to Maintain Computerized Procedure System in the same Version

International Nuclear Information System (INIS)

Jung, Yeon Sub

2016-01-01

CPS(Computerized Procedure System) has been introduced in advanced nuclear power plant worldwide. Korean CPS complies with this trend. It has been applied to Shinkori3,4 and all their successors including BNPP1,2,3,4. CPS is designed by KHNP CRI, and developed by Westinghouse or Doosan. Initial design has been improved by reflecting operator experience, human factor V and V, and advanced engineering method. Design improvement makes CPS differ from plant to plant. For example CPS in Shinhanul cannot interpret CP(Computerized Procedure) of ShinKori3, 4. Different CPS makes it difficult to utilize CP and train operators. Therefore it is important to keep CPS in single version. Note that typical text editors in the office are kept in single version. On the other hand, plant monitoring systems in main control room are actually different. Because CPS is positioned between text editor and monitoring system, goal to maintain CPS in single version is not an easy task. From Shinkori3,4, lots of improvements are achieved in CPS. This improvement surely makes operation easy and will not introduce human errors. Side effect of improvement is that PXS and CPS ES may not be compatible among nuclear power plants. So the side effect should be avoided by minimizing change of XML. Initial XML had been developed for Shinkori3,4. The XML has been modified slightly for Shinhanul1,2. PXS and CPS ES are being developed for the same version for all nuclear power plants of APR1400

Simplified study for the proposed APR1400 condenser performance based on EL-DABAA SITE, EGYPT, with respect to BNPP and SHIN KORI NPP

Energy Technology Data Exchange (ETDEWEB)

Abdoelatef, M. Gomaa; Rezk, Osama; Lee, YonKwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

• Perception of nuclear power as a convenient, economically competitive and viable source of energy which, if introduced in the country, would not only complement the traditional energy sources, but would also promote technological development and serve as an incentive for social and economic progress. In this paper we will introduce APR-1400 as the most likely selected reactor type for 1st Egyptian Nuclear Power Plant. Then we will study the optimized APR-1400 condenser performance based on El-Dabaa site conditions. As the climate conditions, especially the seawater temperatures (intake for the cooling system) having a direct effects on the condenser performance and the overall NPP Efficiency, we studied the seawater temperature change based on EL-Dabaa site conditions and it's feedback on the proposed APR-1400 condenser performance and the NPP efficiency. All of the calculations and analysis had been executed by using KINGS NPA by applying EL-Dabaa seawater temperature variations for the 2 types of APR-1400 Condensers (Shinkori, south Korea and Barakah, UAE), and the main findings were as follow: For the APR-1400 condensers (Shinkori and Barakah) performance • The average difference all over the year for saturation temperature, cooling water outlet temperature, and condenser pressure are 0.135 .deg. C, .05785 .deg. C, and 0.3948 mHgA respectively. This shows that the Barakah condenser performance is not significantly better than the shinkori condenser performance based on EL-Dabaa Conditions, especially if we considered the cost difference between the two condensers designs. For APR-1400 NPPs (Shinkori and Barakah) net efficiency • The average difference all over the year for the power productions and net efficiency are 1.4791 MWe and .00037 %, respectively.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Ualikhan Zhiyenbayev [KAIST, Daejeon (Korea, Republic of); Chung, Dae Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

International Nuclear Information System (INIS)

Ualikhan Zhiyenbayev; Chung, Dae Wook

2015-01-01

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether

Analysis of Turbine Load Rejection for APR1400 using SPACE

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Jin; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

Turbine Load Rejection event is one of the Performance Related Design Basis Event (PRDBE) that can be stabilized using plant control systems without any safety system actuation. The initiation of the event is turbine load rejection from 100% to 5% in 0.019 seconds. The NSSS control systems of APR1400 is composed of the Power Control System (PCS) and the Process-Component Control System (P-CCS). The PCS includes Reactor Regulating System (RRS), Reactor Power Cutback System (RPCS) and Digital Rod Control System (DRCS). The P-CCS includes the Pressurizer Pressure Control System (PPCS), the Pressurizer Level Control System (PLCS), the Feedwater Control System (FWCS) and the Steam Bypass Control System (SBCS). Turbine load rejection results in the increase of secondary pressure due to sudden blocking of steam flow to turbine. Then the Reactor Coolant System (RCS) cooling through steam generators is decreased rapidly and the RCS temperature will be increased. Turbine load rejection is a typical event to test NSSS control systems since it requires the automatic response of all major NSSS control systems. It is shown that the NSSS control systems of APR1400 have the capability to stabilize the plant without any safety system actuation for turbine load rejection event. This analysis results show that SPACE code has the capability to analyze the turbine load rejection event. However, further validation is necessary for other PRDBEs such as Two Main Feedwater Pumps Trip, Turbine Load Step Change and Turbine Load Ramp Down (5%/min) to verify the capability of SPACE for the full range of performance analyses.

Analysis of Turbine Load Rejection for APR1400 using SPACE

International Nuclear Information System (INIS)

Kim, Sang Jin; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon

2016-01-01

Turbine Load Rejection event is one of the Performance Related Design Basis Event (PRDBE) that can be stabilized using plant control systems without any safety system actuation. The initiation of the event is turbine load rejection from 100% to 5% in 0.019 seconds. The NSSS control systems of APR1400 is composed of the Power Control System (PCS) and the Process-Component Control System (P-CCS). The PCS includes Reactor Regulating System (RRS), Reactor Power Cutback System (RPCS) and Digital Rod Control System (DRCS). The P-CCS includes the Pressurizer Pressure Control System (PPCS), the Pressurizer Level Control System (PLCS), the Feedwater Control System (FWCS) and the Steam Bypass Control System (SBCS). Turbine load rejection results in the increase of secondary pressure due to sudden blocking of steam flow to turbine. Then the Reactor Coolant System (RCS) cooling through steam generators is decreased rapidly and the RCS temperature will be increased. Turbine load rejection is a typical event to test NSSS control systems since it requires the automatic response of all major NSSS control systems. It is shown that the NSSS control systems of APR1400 have the capability to stabilize the plant without any safety system actuation for turbine load rejection event. This analysis results show that SPACE code has the capability to analyze the turbine load rejection event. However, further validation is necessary for other PRDBEs such as Two Main Feedwater Pumps Trip, Turbine Load Step Change and Turbine Load Ramp Down (5%/min) to verify the capability of SPACE for the full range of performance analyses

Constructibility assessment of APR1400

International Nuclear Information System (INIS)

Cho, Sung Jae; Kang, Yong Chul; Lee, Jae Gon; Lim, Woo Sang

2003-01-01

APR1400 (formerly KNGR) development is one of national G-7 projects sponsored by government and KHNP (Korea Hydro and Nuclear Power Co., LTD). It is an evolutionary standard reactor with the thermal output of 4000MWt and has been adopted for Shin-Kori Unit 3 and 4. The government organizations, research institutes, universities and industries have participated in the project since 1992 along with KHNP, and the standard design certification program was issued in May 2002. The project set up the top-tier requirements about the safety, economy, operability and maintainability, and constructibility in the early design stage. The requirements have been evaluated periodically during the design process, and the results were reflected to the design. This paper describes the methods for constructibility enhancement and the results of schedule analysis to assure meeting construction duration target, which is set to 48 months from the first concrete pouring to the commercial operation on the condition that learning effects are maximized at Nth plant. To meet the target schedule, the design characteristics and constructibility studies such as new construction methods and construction schedule analysis were performed. The new construction methods presented here are over the top method for NSSS components, deck plate and steel from for concrete wall and slab, automatic welding for large bore piping, and modularization of components and structure, etc. (author)

Evaluation Methodology for Void Swelling Susceptibility of APR1400 Reactor Vessel Internals for U.S. NRC Design Certification

Energy Technology Data Exchange (ETDEWEB)

Kweon, Hyeong Do; Lee, Do Hwan [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The APR1400 RVI (Reactor Vessel Internals) operates in harsh conditions, such as long term exposure to neutron irradiation, high temperatures, reactor coolant environment, and other operating loads. Therefore, even though the RVI components are mainly made of austenitic stainless steel which is well known to have good mechanical and corrosion-resistive properties, these operating conditions. The aging is characterized by a chromium depletion along grain boundaries of austenitic stainless steel, a decrease in ductility and fracture toughness of the steel, an increase in yield and ultimate strength of the steel, and a potential volume change due to void formation in the steel. For these reasons, under certain conditions of stress, temperature, and level of irradiation, the void swelling which is one of the challenging degradation mechanisms affecting the integrity of the RVI may appear at specific locations of the RVI, especially due to high neutron fluence and high temperature under localized gamma heating and low velocity of coolant flow. To assess the effects of operating neutron fluences, temperatures and stresses on the material properties changes and the susceptibility to the void swelling, the evaluation methodology of the APR1400 RVI components for U.S. NRC Design Certification was suggested in this paper. The approach to the evaluation is summarized as follows: 1. RVI component list of the APR1400 is collected. 2. Initial screening to determine the evaluation scope is completed using the design values of fluences. 3. Functionality assessments (radiation transport analysis, CFD analysis, structural analysis) are sequentially performed. 4. Susceptibility to the void swelling is identified through ANSYS/USERMAT module. KHNP believes that the proposed methodology which is based on the EPRI works for operating reactors is the best way to evaluate the void swelling for new reactors such as the APR1400.

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.

Analysis of MSGTR events for APR1400 by means of best estimate thermal-hydraulic system code

International Nuclear Information System (INIS)

Jeong, Ji Hwan; Kim, Sang Jae; Chang, Keun Sun; Lee, Jae Hun

2001-01-01

A multiple steam generator tube rupture (MSGTR) event has never occurred in the history of commercial nuclear reactor operation while single steam generator tube rupture (SGTR) event is reported to occur every two years. As there is no history of MSGTR event, the understandings of transients and consequences of this event are not so much. In this study, a postulated MSGTR event in advanced power reactor 1400 (APR1400) is analyzed using thermal-hydraulic system code. The APR 1400 is a two-loop, 1000 MWe, PWR supposed to be built in 2009. MARS1.4 is used in this study. The present study aims to understand the effects of rupture location in heat transfer tubes and selection of affected steam generator following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 is to allow shortest time for operator action following a tubes rupture in the vicinity of hot-leg side tube sheet and to allow longest time following a tube ruptures at the tube top. The MSSV lift time for rupture at tube-top is evaluated as 24.5% larger than that for rupture at hot-leg side tube sheet. Also, the MSSV lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generator is affected. The comparison shows that the cases for both of two steam generators are affected allow longer time for operator action compared with the cases that a single steam generator is affected. Further more, the tube ruptures in the steam generator where a pressurizer is linked leads to the shortest operator response time

A Development Method of Mobile Computerized Procedure System for the Cooperation among Field Workers and Main Control Room Operators in Korean Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Sun Jin; Seong, No Kyu; Jung, Yeon Sub

2014-01-01

Human errors can occur during the test and maintenance of steam generator, safety injection system and other various systems and devices in nuclear power plants (NPPs). Most of human errors can be improved by the human error prevention techniques such as self-check, peer-check, concurrent verification and etc. Another important technique is to share work information among main control room (MCR) operators and field workers. Various field service automation tools have been developed with recent information technology in many countries. APR1400 computerized procedure system (CPS) has been developed for the MCR operators of Shin-Kori 3 and 4 units. Especially, the concurrent verification support design is applied in the construction project of Shin-Hanul 1 and 2 CPS. It is expected that the proposed mobile CPS can enhance the reduction of human errors by supporting human error prevention techniques and information sharing. This paper describes the technical issues of the mobile CPS (mobile CPS) in the initial development stage. Based on the design of APR1400, CRI CPS has been developed and operated for SKN 3 and 4 HFE V and V and license test for the MCR operating staff. Therefore the mobile CPS will be developed by upgrading the CRI CPS with improved features

A Development Method of Mobile Computerized Procedure System for the Cooperation among Field Workers and Main Control Room Operators in Korean Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Sun Jin; Seong, No Kyu; Jung, Yeon Sub [KHNP ,Daejeon (Korea, Republic of)

2014-08-15

Human errors can occur during the test and maintenance of steam generator, safety injection system and other various systems and devices in nuclear power plants (NPPs). Most of human errors can be improved by the human error prevention techniques such as self-check, peer-check, concurrent verification and etc. Another important technique is to share work information among main control room (MCR) operators and field workers. Various field service automation tools have been developed with recent information technology in many countries. APR1400 computerized procedure system (CPS) has been developed for the MCR operators of Shin-Kori 3 and 4 units. Especially, the concurrent verification support design is applied in the construction project of Shin-Hanul 1 and 2 CPS. It is expected that the proposed mobile CPS can enhance the reduction of human errors by supporting human error prevention techniques and information sharing. This paper describes the technical issues of the mobile CPS (mobile CPS) in the initial development stage. Based on the design of APR1400, CRI CPS has been developed and operated for SKN 3 and 4 HFE V and V and license test for the MCR operating staff. Therefore the mobile CPS will be developed by upgrading the CRI CPS with improved features.

Probability of Neutralization Estimation for APR1400 Physical Protection System Design Effectiveness Evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Myungsu; Lim, Heoksoon; Na, Janghwan; Chi, Moongoo [Korea Hydro and Nuclear Power Co. Ltd. Central Research Institute, Daejeon (Korea, Republic of)

2015-05-15

It is focusing on development of a new designing process which can be compatible to international standards such as IAEA1 and NRC2 suggest. Evaluation for the design effectiveness was found as one of the areas to improve. If a design doesn't meet a certain level of effectiveness, it should be re-designed accordingly. The effectiveness can be calculated with combination of probability of Interruption and probability of neutralization. System Analysis of Vulnerability to Intrusion (SAVI) has been developed by Sandia National Laboratories for that purpose. With SNL's timely detection methodology, SAVI has been used by U.S. nuclear utilities to meet the NRC requirements for PPS design effectiveness evaluation. For the SAVI calculation, probability of neutralization is a vital input element that must be supplied. This paper describes the elements to consider for neutralization, probability estimation methodology, and the estimation for APR1400 PPS design effectiveness evaluation process. Markov chain and Monte Carlo simulation are often used for simple numerical calculation to estimate P{sub N}. The results from both methods are not always identical even for the same situation. P{sub N} values for APR1400 evaluation were calculated based on Markov chain method and modified to be applicable for guards/adversaries ratio based analysis.

Measurement techniques of local parameters in the downcomer boiling experiment of APR1400

International Nuclear Information System (INIS)

Lee, Eu Hwak

2004-02-01

In order to investigate boiling phenomena experimentally in the downcomer during LBLOCA with Direct Vessel Injection (DVI), which is a new Safety Injection System (SIS) of Advanced Power Reactor 1400 MW (APR1400), several parameters should be measured through the verification of their applicability. In this study, measurement techniques of the parameters are developed for the downcomer boiling experiment; local phase velocities, local void fraction and heat flux from the heated wall. The experiment has been performed with the heated wall, which has a thickness of 8.2 cm and a height of 32.5 cm and made of the same material as the prototype (APR1400) with chrome coating against rusting. The newly developed pitot tube is applied to the measurement of local liquid velocity and its calibration curve is obtained experimentally with the consideration of the effect according to water temperature and hole size changes. The developed pitot tube measures the local liquid velocity with 0.69 % deviation and it is confirmed that the water temperature and geometrical change does not affect the calibration curve. The high-speed camera and commercial software are used to measure the local vapor velocity with the accuracy of 0.06 m/sec per pixel and the procedure is confirmed in the present study. It turns out that the vapor velocity is insensitive to void size. High-speed camera and image processing are used to measure the local void fraction with the determined intensity criterion for distinguishing each phase and the results are compared with the bulk void fraction by differential pressure transmitters. In the actual experiment, the developed method is applied successfully and the results show that the criterion of intensity has little effect on local void fraction. And, it is observed that the tendency between the measured local and bulk void faction is maintained with time. In order to measure heat flux from the heated wall, two heat flux measurement techniques are developed

Domestic activity for technical development of the APR1400's RCP performance test

Energy Technology Data Exchange (ETDEWEB)

Cho, Seok; Kim, Seok; Bae, Byung-Uhn; Cho, Yun-Jae; Kim, Yeon-Sik; Jeon, Woo-Jin; Yun, Young-Jung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The thermal hydraulic and electric capability of the RCP test facility (RCPTF) covers up to 18.5 MPa, 343 .deg. C, 11.7 m{sup 3}/s, and 14.0 MW in the design pressure, temperature, flow rate, and the maximum electric power, respectively. In 2013, commissioning test had been performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance verification and evaluation capability of the RCP performance. After finishing the commissioning and modification of the RCPTF, type test for the new-type RCP had been performed successfully. In this paper, several technical issues developed in the 2013 and the type test's method and results will be described. In the present paper, the technical activities for the development of the verification test of APR1400's RCP are described. KAERI has completed the full set of technology development, prerequisite for the RCP verification test, and now on the way to perform a test for the sealing capacity of the seal assembly during the Station Block Out (SBO) condition of APR1400.

CFD Analysis of Random Turbulent Flow Load in Steam Generator of APR1400 Under Normal Operation Condition

International Nuclear Information System (INIS)

Lim, Sang Gyu; You, Sung Chang; Kim, Han Gon

2011-01-01

Regulatory guide 1.20 revision 3 of the Nuclear Regulatory Committee (NRC) modifies guidance for vibration assessments of reactor internals and steam generator internals. The new guidance requires applicants to provide a preliminary analysis and evaluation of the design and performance of a facility, including the safety margins of during normal operation and transient conditions anticipated during the life of the facility. Especially, revision 3 require rigorous assessments of adverse flow effects in the steam dryer cased by flow-excited acoustic and structural resonances such as the abnormality from power-uprated BWR cases. For two nearly identical nuclear power plants, the steam system of one BWR plant experienced failure of steam dryers and the main steam system components when steam flow was increased by 16 percent for extended power uprate (EPU). The mechanisms of those failures have revealed that a small adverse flow changing from the prototype condition induced severe flow-excited acoustic and structural resonances, leading to structural failures. In accordance with the historical background, therefore, potential adverse flow effects should be evaluated rigorously for steam generator internals in both BWR and Pressurized Water Reactor (PWR). The Advanced Power Reactor 1400 (APR1400), an evolutionary light water reactor, increased the power by 7.7 percent from the design of the 'Valid Prototype', System80+. Thus, reliable evaluations of potential adverse flow effects on the steam generator of APR1400 are necessary according to the regulatory guide. This paper is part of the computational fluid dynamics (CFD) analysis results for evaluation of the adverse flow effect for the steam generator internals of APR1400, including a series of sensitivity analyses to enhance the reliability of CFD analysis and an estimation the effect of flow loads on the internals of the steam generator under normal operation conditions

Application of the risk-informed methodology for APR1400 P-T limits curve

Energy Technology Data Exchange (ETDEWEB)

Kim, K.; Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-07-01

A reactor pressure vessel (RPV) in a nuclear power plant has operational limits of pressure and temperature to prevent a potential drastic propagation of cracks due to brittle fracture. We call it a pressure-temperature limits curve (P-T limits curve). Appendix G of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, provides deterministic procedures to develop the P-T limits curve for the reactor pressure vessel. Recently, an alternative risk-informed methodology has been added in the ASME Code. Risk-informed means that we can consider insights from a probabilistic risk assessment by using this methodology. This alternative methodology provides a simple procedure to develop risk-informed P-T limits for heat up, cool down, and hydrostatic test events. The risk-informed P-T limits curve is known to provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. In this paper, we developed both the deterministic and a risk-informed P-T limits curve for an APR1400 reactor using Appendix G of the ASME Code, Section XI and compare the results in terms of additional operational margin. (author)

Response Time Analysis and Test of Protection System Instrument Channels for APR1400 and OPR1000

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang Jae; Han, Seung; Yun, Jae Hee; Baek, Seung Min [Department of Instrumentation and Control System Engineering, KEPCO Engineering and Construction, Daejeon (Korea, Republic of); Lee, Sang Jeong [Department of Electronics Engineering, Chungnam National University, Daejeon (Korea, Republic of)

2015-07-01

Safety limits are required to maintain the integrity of physical barriers designed to prevent the uncontrolled release of radioactive materials in nuclear power plants. The safety analysis establishes two critical constraints that include an analytical limit in terms of a measured or calculated variable, and a specific time after the analytical limit is reached to begin protective action. Keeping with the nuclear regulations and industry standards, satisfying these two requirements will ensure that the safety limit will not be exceeded during the design basis event, either an anticipated operational occurrence or a postulated accident. Various studies on the setpoint determination methodology for the safety-related instrumentation have been actively performed to ensure that the requirement of the analytical limit is satisfied. In particular, the protection setpoint methodology for the advanced power reactor 1400 (APP1400) and the optimized power reactor 1000 (OPR1000) has been recently developed to cover both the design basis event and the beyond design basis event. The developed setpoint methodology has also been quantitatively validated using specific computer programs and setpoint calculations. However, the safety of nuclear power plants cannot be fully guaranteed by satisfying the requirement of the analytical limit. In spite of the response time verification requirements of nuclear regulations and industry standards, it is hard to find the studies on the systematically integrated methodology regarding the response time evaluation. In cases of APR1400 and OPR1000, the response time analysis for the plant protection system is partially included in the setpoint calculation and the response time test is separately performed via the specific plant procedure. The test technique has a drawback which is the difficulty to demonstrate completeness of timing test. The analysis technique has also a demerit of resulting in extreme times that not actually possible. Thus

Response Time Analysis and Test of Protection System Instrument Channels for APR1400 and OPR1000

International Nuclear Information System (INIS)

Lee, Chang Jae; Han, Seung; Yun, Jae Hee; Baek, Seung Min; Lee, Sang Jeong

2015-01-01

Safety limits are required to maintain the integrity of physical barriers designed to prevent the uncontrolled release of radioactive materials in nuclear power plants. The safety analysis establishes two critical constraints that include an analytical limit in terms of a measured or calculated variable, and a specific time after the analytical limit is reached to begin protective action. Keeping with the nuclear regulations and industry standards, satisfying these two requirements will ensure that the safety limit will not be exceeded during the design basis event, either an anticipated operational occurrence or a postulated accident. Various studies on the setpoint determination methodology for the safety-related instrumentation have been actively performed to ensure that the requirement of the analytical limit is satisfied. In particular, the protection setpoint methodology for the advanced power reactor 1400 (APP1400) and the optimized power reactor 1000 (OPR1000) has been recently developed to cover both the design basis event and the beyond design basis event. The developed setpoint methodology has also been quantitatively validated using specific computer programs and setpoint calculations. However, the safety of nuclear power plants cannot be fully guaranteed by satisfying the requirement of the analytical limit. In spite of the response time verification requirements of nuclear regulations and industry standards, it is hard to find the studies on the systematically integrated methodology regarding the response time evaluation. In cases of APR1400 and OPR1000, the response time analysis for the plant protection system is partially included in the setpoint calculation and the response time test is separately performed via the specific plant procedure. The test technique has a drawback which is the difficulty to demonstrate completeness of timing test. The analysis technique has also a demerit of resulting in extreme times that not actually possible. Thus

Numerical Investigation of APR1400 Loop Seal Reformation and Clearing Phenomena during Post-LOCA Long-term Cooling Period using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Lee, S. I.; Jeon, J. H.; Park, J. H.; Sung, H. J.; Lee, J. I.; Choi, T. S. [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

Hand calculations had been tried to show that the loop seal reformation does not degrade the LTC performance of the APR1400, but the result was unclear. Therefore, the code assessments of the phenomena for APR1400 are required to guarantee the performance of LTC. In the present study, RELAP5/MOD3.3 patch04 (3.3iy) is used for the assessments. The safety degradation, e.g., possible cladding temperature rise due to LSRC during LTC period would be stronger when the safety injection (SI) is larger, while the core uncovery is more probable when the SI is smaller. In other words, it is not easy to determine the limiting condition of safety injection as two or four safety injection pumps (SIPs) in APR1400. In the present study, both the minimum SI condition with two SIPs and the maximum SI condition with four SIPs are considered. The study focuses only on the top slot break condition due to its highest possibility of LSRC phenomena. The LSRC issues for APR1400 are discussed and investigated through numerical investigation using RELAP5MOD3.3 patch04 for top-slot cold leg (pump discharge line) breaks. The major items investigated are whether the simultaneous four loop seal reformation is possible and whether there is any possible violation of the long term coolability criteria in 10 CFR 50.46. Safety injection with two SIPs and four SIPs are investigated to find that the four loop seal reformation condition before the simultaneous injection may occur only with four SIPs.

Post-test analysis for the APR1400 LBLOCA DVI performance test using MARS

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Lee, Y. J.; Kim, H. C.; Bae, Y. Y.; Park, J. K.; Lee, W.

2002-03-01

Post-test analyses using a multi-dimensional best-estimate analysis code, MARS, are performed for the APR1400 LBLOCA DVI (Direct Vessel Injection) performance tests. This report describes the code evaluation results for the test data of various void height tests and direct bypass tests that have been performed at MIDAS test facility. MIDAS is a scaled test facility of APR1400 with the objective of identifying multi-dimensional thermal-hydraulic phenomena in the downcomer during the reflood conditions of a large break LOCA. A modified linear scale ratio was applied in its construction and test conditions. The major thermal-hydraulic parameters such as ECC bypass fraction, steam condensation fraction, and temperature distributions in downcomer are compared and evaluated. The evaluation results of MARS code for the various test cases show that: (a) MARS code has an advanced modeling capability of well predicting major multi-dimensional thermal-hydraulic phenomena occurring in the downcomer, (b) MARS code under-predicts the steam condensation rates, which in turn causes to over-predict the ECC bypass rates. However, the trend of decrease in steam condensation rate and increase in ECC bypass rate in accordance with the increase in steam flow rate, and the calculation results of the ECC bypass rates under the EM analysis conditions generally agree with the test data

The Analysis of Uncontrolled CEA Bank withdrawal from low power Accident for APR1400 Using KNAP

International Nuclear Information System (INIS)

Yang, Chang-Keun; Kim, Yo-Han; Ha, Sang-Jun

2006-01-01

In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering(CE), for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may cause difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor (PWR) has been raised. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the review on several candidates, RETRAN-3D has been chosen as a system analysis code. The draft version of the methodology was developed based on the references for the general purpose, and modified to apply it to specific plants in Korea. As a part of the feasibility estimation for the methodology and code system, uncontrolled CEA bank withdrawal from low power accident for the Advanced Power Reactor 1400(APR1400) was selected to verify the feasibility using the RETRAN-3D. And the results were compared with the Standard Safety Analysis Reports (SSAR) of APR1400

Analysis Of Feedwater Line Break Of APR1400 By MARS Code

International Nuclear Information System (INIS)

Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

2011-01-01

This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

Analysis Of Control Rod Ejection Of APR1400 By RELAP5

International Nuclear Information System (INIS)

Le Thi Thu; Hoang Minh Giang; Vo Thi Huong; Le Dai Dien

2011-01-01

This paper presents the analysis of Reactivity Induced Accident caused by ejection of a Control Element Assembly (CEA) from APR 1400 reactor vessel within 0.05 second. The initial condition were assumed as following: power level at 102%, delayed neutron fraction β = 412 pcm and CEA worth = 110 pcm. The analysis was simulated by RELAP5 code through two step: calculation of steady state and calculation of transient with initial condition mentioned as above. Some output results were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as power, reactivity feedback from fuel temperature changes (Doppler) as well as temperature, pressure, DNBR within 6 second of the accident. (author)

Preliminary design of S-CO{sub 2} Brayton cycle for APR-1400 with power generation and desalination process

Energy Technology Data Exchange (ETDEWEB)

Bae, Seong Jun; Lee, Won Woong; Jeong, Yong Hoon; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

2015-10-15

This study was conducted to explore the capabilities of the S-CO{sub 2} Brayton cycle for a cogeneration system for APR-1400 application. Three concepts of the S-CO{sub 2} simple recuperated co-generation cycle were designed. A supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle is recently receiving significant attention as a promising power conversion system in wide range of energy applications due to its high efficiency and compact footprint. The main reason why the S-CO{sub 2} Brayton cycle has these advantages is that the compressor operates near the critical point of CO{sub 2} (30.98 .deg. C, 7.38MPa) to reduce the compression work significantly compared to the other Brayton cycles. In this study, the concept of replacing the entire steam cycle of APR-1400 with the S-CO{sub 2} Brayton cycle is evaluated. The power generation purpose S-CO{sub 2} Brayton cycles are redesigned to generate power and provide heat to the desalination system at the same time. The performance of these newly suggested cycles are evaluated in this paper. The target was to deliver 147MW heat to the desalination process. The thermal efficiencies of the three concepts are not significantly different, but the 3{sup rd} concept is relatively simpler than other cycles because only an additional heat exchanger is required. Although the 2{sup nd} concept is relatively complicated in comparison to other concepts, the temperatures at the inlet and outlet of the DHX are higher than that of the others. As shown in the results, the S-CO{sub 2} Brayton cycles are not easy to outperform the steam cycle with very simple layout and general design points under APR-1400 operating condition. However, this study shows that the S-CO{sub 2} Brayton cycles can be designed as a co-generation cycle while producing the target desalination heat with a simple configuration. In addition, it was also found that the S-CO{sub 2} Brayton cycle can achieve higher cycle thermal efficiency than the steam power cycle under

Best-estimated multi-dimensional calculation during LB LOCA for APR1400

International Nuclear Information System (INIS)

Oh, D. Y.; Bang, Y. S.; Cheong, A. J.; Woong, S.; Korea, W.

2010-01-01

Best-estimated (BE) calculation with uncertainty quantification for the emergency core cooling system (ECCS) performance analysis during Loss of Coolant Accident (LOCA) is more broadly used in nuclear industries and regulations. In Korea, demand on regulatory audit calculation is continuously increasing to support the safety review for life extension, power up-rating and advanced nuclear reactor design. The thermal-hydraulic system code, MARS (Multi-dimensional Analysis of Reactor Safety), with multi-dimensional capability is used for audit calculation. It achieves to describe the complicated phenomena in reactor coolant system by very effectively consolidating the one dimensional RELAP5/MOD3 with the multidimensional COBRA-TF codes. The advanced power reactors (APR1400) to be evaluated has four separated hydraulic trains of the high pressure injection system (HPSI) with direct vessel injection (DVI) which is different from the existing commercial PWRs. Also, the therma-hydraulic behavior of DVI plant would be considerably different from that of a cold-leg safety injection since the low pressure safety injection system are eliminated and the high pressure safety flow are injected into the specific elevation of reactor vessel downcomer. The ECCS bypass induced by the downcomer boiling due to hot wall heating of reactor vessel during reflooding phase is one of the important phenomena which should be considered in DVI plants. Therefore, in this study, BE calculation with one-dimensional (1-D) and multi-dimensional (multi-D) MARS models during LBLOCA are performed for APR1400 plant. In the multi-D evaluation, the reactor vessel is modeled by multi-D components and the specific treatment of flow path inside reactor vessel, e.g., upper guide structure, is essential. The concept of hot zone is adopted to simulate the limiting thermal-hydraulic conditions surrounding hot rod, which is similar to hot channel in 1-D. Also, alternative treatment of the hot rods in multi-D is

Comparison of APR1400 safety between brake site and shin-Kori site Due to the difference in the climate conditions

International Nuclear Information System (INIS)

Yoon, Ho Joon; Lee, Jeong Ik; Lee, Jeong Ik

2012-01-01

Brake Nuclear Power Plant (BNPP) is now under the construction based on APR1400 designed by Korean Electric Power Corporation (KEPCO). APR1400 is a two loop pressurized water reactor, the nuclear steam supply system (NSSS) US designed for about put of 4,000 MWt, with a corresponding electrical output of approximately 1,390 MWe. The first APR1400 (SKN 3 and 4) constructed in Shin-Kori, Korea has been modified according to the surrounding environment of the United Arab Emirates. In this paper, authors would like to compare safety issues between B NPP and Skin due to the changes of surroundings, since the site characteristics are very different. For instance, the mean annual air temperature in the UAE is 28 .deg. C and the peak air temperature was recorded as 48.8 .deg. C. Sea temperatures are varying from 17. deg. C in January to 35. deg. C in August, while that of Korea is in 9-16. deg. C range. Hot climate of UAE and the malfunction of HVAC system can lead the increasing of the water temperature in safety injection system (SIS). The heated water in SIS may affect the safety margin of the peak cladding temperature (PCT). The change of PCT and response time according to design basis accident scenarios such as large break LOCA are analyzed in detail. To evaluate such effect, Mars code was utilized to evaluate assumed condition by KAIST and the analyses of the results were carried out by Khalifa Univ.

Reliability analysis of protection system of advanced pressurized water reactor - APR 1400

International Nuclear Information System (INIS)

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

2003-04-01

Reliability analysis was carried out for the protection system of the Korean Advanced Pressurized Water Reactor - APR 1400. The main focus of this study was the reliability analysis of digital protection system, however, towards giving an integrated statement of complete protection reliability an attempt has been made to include the shutdown devices and other related aspects based on the information available to date. The sensitivity analysis has been carried out for the critical components / functions in the system. Other aspects like importance analysis and human error reliability for the critical human actions form part of this work. The framework provided by this study and the results obtained shows that this analysis has potential to be utilized as part of risk informed approach for future design / regulatory applications

Preliminary Analysis of Ex-Vessel Steam Explosion using TEXAS-V code for APR1400

International Nuclear Information System (INIS)

Song, Sung Chu; Lee, Jung Jae; Cho, Yong Jin; Hwang, Taesuk

2013-01-01

The purpose of this study is to explore input development and the audit calculation using TEXAS-V code for ex-vessel steam explosion for a flooded reactor cavity of APR1400. TEXAS computational models are one of the simplified tools for simulations of fuel-coolant interaction during mixing, triggering and explosion phase. The models of TEXAS code were validated by performing the fundamental experimental investigation in the KROTOS facility at JRC, Ispra. The experiments such as KROTOS and FARO experiment are aimed at providing benchmark data to examine the effect of fuel-coolant initial conditions and mixing on explosion energetics with alumina and prototypical core material. TEXAS-V code used in this study was to analyze and predict the ex-vessel steam explosion for a reactor scale. The input deck to simulate the flooded reactor cavity of APR1400 is developed and base case calculation is performed. This study will provide a base for further study. The code will be of use for the evaluation and sensitivity study of ex-vessel steam explosion for ERVC strategy in the future studies. Analysis result of this study is similar to the result of other study. Through this study, it is found that TEXAS-V could be the used as a tool for predicting the steam explosion load on a reactor scale, as fast running computer code. In addition, TEXAS-V code could be to evaluate the impact of various uncertainties, which are not clearly understood yet, to provide a conservative envelope for the steam explosion

Preliminary Analysis of Ex-Vessel Steam Explosion using TEXAS-V code for APR1400

Energy Technology Data Exchange (ETDEWEB)

Song, Sung Chu; Lee, Jung Jae; Cho, Yong Jin; Hwang, Taesuk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

The purpose of this study is to explore input development and the audit calculation using TEXAS-V code for ex-vessel steam explosion for a flooded reactor cavity of APR1400. TEXAS computational models are one of the simplified tools for simulations of fuel-coolant interaction during mixing, triggering and explosion phase. The models of TEXAS code were validated by performing the fundamental experimental investigation in the KROTOS facility at JRC, Ispra. The experiments such as KROTOS and FARO experiment are aimed at providing benchmark data to examine the effect of fuel-coolant initial conditions and mixing on explosion energetics with alumina and prototypical core material. TEXAS-V code used in this study was to analyze and predict the ex-vessel steam explosion for a reactor scale. The input deck to simulate the flooded reactor cavity of APR1400 is developed and base case calculation is performed. This study will provide a base for further study. The code will be of use for the evaluation and sensitivity study of ex-vessel steam explosion for ERVC strategy in the future studies. Analysis result of this study is similar to the result of other study. Through this study, it is found that TEXAS-V could be the used as a tool for predicting the steam explosion load on a reactor scale, as fast running computer code. In addition, TEXAS-V code could be to evaluate the impact of various uncertainties, which are not clearly understood yet, to provide a conservative envelope for the steam explosion.

Minimum containment pressure and its effect on ECCS performance of APR-1400

International Nuclear Information System (INIS)

Kim, In Goo; Bang, Young S.; Kim, Hho Jung

2004-01-01

The containment pressure has a strong effect on the late reheat behavior for a large break LOCA, associated with the DVI issue. The downcomer boiling, which occurs during the post-reflood phase, has a negative effect on core cooling for a LBLOCA. Because the downcomer boiling is enhanced as the containment pressure decreases, how to determine containment pressure is important to the evaluation of ECCS performance. In spite of its importance of containment pressure, there are few studies on the containment pressure and the interaction between RCS and containment thermal hydraulics. To have a better knowledge of the effect of containment pressure on APR-1400 ECCS performance, a parametric study for containment pressure has been carried out. Also, the interaction between RCS and containment behavior has been also investigated

Use case driven approach to develop simulation model for PCS of APR1400 simulator

International Nuclear Information System (INIS)

Dong Wook, Kim; Hong Soo, Kim; Hyeon Tae, Kang; Byung Hwan, Bae

2006-01-01

The full-scope simulator is being developed to evaluate specific design feature and to support the iterative design and validation in the Man-Machine Interface System (MMIS) design of Advanced Power Reactor (APR) 1400. The simulator consists of process model, control logic model, and MMI for the APR1400 as well as the Power Control System (PCS). In this paper, a use case driven approach is proposed to develop a simulation model for PCS. In this approach, a system is considered from the point of view of its users. User's view of the system is based on interactions with the system and the resultant responses. In use case driven approach, we initially consider the system as a black box and look at its interactions with the users. From these interactions, use cases of the system are identified. Then the system is modeled using these use cases as functions. Lower levels expand the functionalities of each of these use cases. Hence, starting from the topmost level view of the system, we proceeded down to the lowest level (the internal view of the system). The model of the system thus developed is use case driven. This paper will introduce the functionality of the PCS simulation model, including a requirement analysis based on use case and the validation result of development of PCS model. The PCS simulation model using use case will be first used during the full-scope simulator development for nuclear power plant and will be supplied to Shin-Kori 3 and 4 plant. The use case based simulation model development can be useful for the design and implementation of simulation models. (authors)

Development of a Human Performance Evaluation Support System for Human Factors Validation of MCR MMI Design in APR-1400

International Nuclear Information System (INIS)

Ha, Jun Su; Seong, Poong Hyun

2005-01-01

As CRT-based display and advanced information technology were applied to advanced reactors such as APR-1400 (Advanced Power Reactor-1400), human operators' tasks became more cognitive works. As a results, Human Factors Engineering (HFE) became more important in designing the MCR (Main Control Room) MMI (Man-Machine Interface) of an advanced reactor. According to the Human Factors Engineering Program Review Model, human factors validation of MCR MMI design should be performed through performance-based tests to determine whether it acceptably supports safe operation of the plant. In order to support the evaluation of the performance, a HUman Performance Evaluation Support System (HUPESS) is in development

Human Reliability Analysis For Computerized Procedures

International Nuclear Information System (INIS)

Boring, Ronald L.; Gertman, David I.; Le Blanc, Katya

2011-01-01

This paper provides a characterization of human reliability analysis (HRA) issues for computerized procedures in nuclear power plant control rooms. It is beyond the scope of this paper to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper provides a review of HRA as applied to traditional paper-based procedures, followed by a discussion of what specific factors should additionally be considered in HRAs for computerized procedures. Performance shaping factors and failure modes unique to computerized procedures are highlighted. Since there is no definitive guide to HRA for paper-based procedures, this paper also serves to clarify the existing guidance on paper-based procedures before delving into the unique aspects of computerized procedures.

A MARS and MIDAS Linked Accident Simulation for Large LOCA in APR1400

International Nuclear Information System (INIS)

Choi, Young; Kim, K. R.; Kim, D. H.; Chung, B. D.

2006-01-01

A linked calculation utilizing the design-basis code MARS and the severe accident code MIDAS has been accomplished for a station blackout simulation in APR1400. The MARS code was developed by using the RELAP3/MOD3 and COBRA-TF codes, while the MIDAS code is currently under a development process using the MELCOR code. The objectives of this paper are to explain how to identify the MAR-MIDAS linked calculation outlines and the technical problems, including the MARS data transfer method, the MIDAS input generation works and so on. For the performance verification of the MARS-MIDAS linked calculation, the MARS, MIDAS and their linkage system are run independently for the same initiating event, so that their data can be compared with each other after the selection of proper variables

Analysis on Containment Response Following a LBLOCA of APR1400

International Nuclear Information System (INIS)

Han, Kyu Hyun

2016-01-01

The predictions are in good agreements with the final safety analysis report, which implies the containment integrity is maintained during or after an accident like loss of coolant accident. In this study, the CONTEMPT-LT/028 was used to calculate the pressure and temperature, and in the follow-up study, CONTAIN 2.0 will be used for the pressure and temperature predictions in APR1400 reactors. Shin-Hanul Units 1 and 2 may possess different characteristics of peak pressure and temperature in containment following a large break loss-of-coolant-accident. To assess the important performance independently and to compare with prediction results presented in the final safety analysis report (FSAR) of Shin-Hanul Units 1 and 2 might be helpful to regulatory review for identifying validity of the FSAR. The end of blowdown (EOB) time during a LOCA could largely affect the peak pressure and temperature in the containment. This paper provides CONTEMPT-LT/028 prediction of the peak pressure and temperature of Shin-Hanul Units 1 and 2 following a large break loss-of-coolant-accident and compares with licensee's prediction results.

Analysis on Containment Response Following a LBLOCA of APR1400

Energy Technology Data Exchange (ETDEWEB)

Han, Kyu Hyun [KINS, Daejeon (Korea, Republic of)

2016-05-15

The predictions are in good agreements with the final safety analysis report, which implies the containment integrity is maintained during or after an accident like loss of coolant accident. In this study, the CONTEMPT-LT/028 was used to calculate the pressure and temperature, and in the follow-up study, CONTAIN 2.0 will be used for the pressure and temperature predictions in APR1400 reactors. Shin-Hanul Units 1 and 2 may possess different characteristics of peak pressure and temperature in containment following a large break loss-of-coolant-accident. To assess the important performance independently and to compare with prediction results presented in the final safety analysis report (FSAR) of Shin-Hanul Units 1 and 2 might be helpful to regulatory review for identifying validity of the FSAR. The end of blowdown (EOB) time during a LOCA could largely affect the peak pressure and temperature in the containment. This paper provides CONTEMPT-LT/028 prediction of the peak pressure and temperature of Shin-Hanul Units 1 and 2 following a large break loss-of-coolant-accident and compares with licensee's prediction results.

Formal Verification of Computerized Procedure with Colored Petri Nets

International Nuclear Information System (INIS)

Kim, Yun Goo; Shin, Yeong Cheol

2008-01-01

Computerized Procedure System (CPS) supports nuclear power plant operators in performing operating procedures which are instructions to guide in monitoring, decision making and controlling nuclear power plants. Computerized Procedure (CP) should be loaded to CPS. Due to its execution characteristic, computerized procedure acts like a software in CPS. For example, procedure flows are determined by operator evaluation and computerized procedure logic which are pre-defined. So the verification of Computerized Procedure logic and execution flow is needed before computerized procedures are installed in the system. Formal verification methods are proposed and the modeling of operating procedures with Coloured Petri Nets(CP-nets) is presented

Aircraft Impact Assessment of APR1400 Reactor Containment Building

International Nuclear Information System (INIS)

Moon, Il Hwan; Kim, Do Yeon; Kim, Jae Hee; Kim, Sang Yun

2011-01-01

The implementation of a protection to withstand aircraft impact on safety-related structures and systems is basically based on a probabilistic evaluation for each site, if the licensing body doesn't require a deterministic approach. Existing nuclear power plants in Korea were designed based on the probabilistic approach, and the aircraft impact hazard remained less than a probability of 10 -7 . However, a man-made aircraft impact have been considered as a possible external accident for the nuclear power plant. New plant designs that are to be constructed in the U.S. after July 2009 must consider the effect of impact from a large commercial aircraft according to the requirements of 10 CFR 50.150. Especially, Reactor Containment Building (RCB) housing the safety-related equipment and fuels should be protected safely against aircraft crash without perforation and scabbing failure of external wall. APR1400 RCB is constructed as a prestressed concrete containment vessel (PCCV) which is surrounded by the auxiliary building housing additional safety-related equipment and other systems. In this study, the aircraft impact analyses for the RCB are carried out using Riera forcing function and aircraft model. Considered external wall thickness is 4 ft 6 in. for the cylindrical wall and 4 ft for the dome. Actual strengths of concrete and steel are considered as the material properties. For these analyses, the dynamic increment factor and concrete aging effect are considered in accordance with NEI 07-13(2011)

Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

International Nuclear Information System (INIS)

Elsawi, Mohamed A.; Alnoamani, Zainab

2015-01-01

The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

Failure Mode and Effect Analysis of the Application Software of the Safety-critical I and C System in APR1400

Energy Technology Data Exchange (ETDEWEB)

Kim, Koheun; Kim, Yong geul; Choi, Woong seok; Sohn, Se do [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

2016-10-15

In APR1400, the computer software hazard analysis is performed by hazard and operability analysis (HAZOP) method. Meanwhile, HAZOP has its limitation and cannot be considered better than fault tree analysis (FTA) or failure mode and effect (FMEA) analysis. HAZOP assumes that the system has been carefully studied, and all possible hazards, their effects or consequences and remedies are incorporated in the system. But incorporating every possible event in the design is impossible. In this light, this paper attempts to use FMEA method for evaluating the risk for safety-critical instrumentation and control (I and C) system software for NPP which is more practically than HAZOP. It is possible because the software failures are due to systematic faults that causing simultaneous failure in multiple division when the triggering event happens. This analysis is applied to safety-critical system of Shin-Hanul units 1 and 2 NPP, i.e., APR1400. Through SFMEA, the critical software failure modes and tasks that could result in CCF are identified and also evaluated to determine the associated risk level (e.g. high or intermediate or low) based on the failure effect. Biggest benefit from this analysis comparing with HAZOP is it can reveal the possible weak points and provide the guidance to the V and V team by helping to generate the test cases.

Analysis of steam condensation in APR1400 IRWST for loss of coolant accident

Energy Technology Data Exchange (ETDEWEB)

Oh, Young Suk

2006-02-15

The In-Containment Refueling Water Storage Tank (IRWST) of APR1400 is installed at the bottom of containment building to promote the plant safety functions during an accident. This design feature brings about uncertainty factors which may necessitate conventional prediction of temperature and pressure of containment building improved or revised when an accident occurs. The hot steam which is released from RCS break enters the IRWST through four Pressure Relief Dampers (PRDs). It is expected to be condensed with water stored in IRWST, in which water is colder than incoming steam. The purpose of this study is to investigate the influence of IRWST and pressure relief damper on back pressure and temperature in APR1400 containment codes such as CONTEMPT-LT and GOTHIC. The comparison of codes showed that GOTHIC code be more appropriate for the prediction of containment pressure and temperature under the condition of steam condensation occurring in confined water pool. Especially, the GOTHIC has superior capability to treat multi-compartmentalized geometry This study developed one-compartment (single) model, two-compartment (separated) model, and three-dimension (3-D) model, respectively. Two compartment model separates the IRWST from the other containment compartments. In 3-D model, only the IRWST is nodalized with Cartesian modeling. The single model is developed for comparison with two-compartment model which can analyze PRD's influence. The separated model for predicting PRD's influence divides the space between containment and IRWST. 3-D model for IRWST was generated because it is not symmetric considering location of sparger, pump, and suction sump. Therefore, IRWST is simulated with not only detailed three-dimensional behavior but also independent flow paths for four PRDs. Many experimental studies for the direct-contact heat transfer in stratified steam water flows, cocurrent or countercurrent, have been performed (Segev et al., 1981; Lim et al., 1981

Analysis of steam condensation in APR1400 IRWST for loss of coolant accident

International Nuclear Information System (INIS)

Oh, Young Suk

2006-02-01

The In-Containment Refueling Water Storage Tank (IRWST) of APR1400 is installed at the bottom of containment building to promote the plant safety functions during an accident. This design feature brings about uncertainty factors which may necessitate conventional prediction of temperature and pressure of containment building improved or revised when an accident occurs. The hot steam which is released from RCS break enters the IRWST through four Pressure Relief Dampers (PRDs). It is expected to be condensed with water stored in IRWST, in which water is colder than incoming steam. The purpose of this study is to investigate the influence of IRWST and pressure relief damper on back pressure and temperature in APR1400 containment codes such as CONTEMPT-LT and GOTHIC. The comparison of codes showed that GOTHIC code be more appropriate for the prediction of containment pressure and temperature under the condition of steam condensation occurring in confined water pool. Especially, the GOTHIC has superior capability to treat multi-compartmentalized geometry This study developed one-compartment (single) model, two-compartment (separated) model, and three-dimension (3-D) model, respectively. Two compartment model separates the IRWST from the other containment compartments. In 3-D model, only the IRWST is nodalized with Cartesian modeling. The single model is developed for comparison with two-compartment model which can analyze PRD's influence. The separated model for predicting PRD's influence divides the space between containment and IRWST. 3-D model for IRWST was generated because it is not symmetric considering location of sparger, pump, and suction sump. Therefore, IRWST is simulated with not only detailed three-dimensional behavior but also independent flow paths for four PRDs. Many experimental studies for the direct-contact heat transfer in stratified steam water flows, cocurrent or countercurrent, have been performed (Segev et al., 1981; Lim et al., 1981; Kim and

Computerized procedures system

Science.gov (United States)

Lipner, Melvin H.; Mundy, Roger A.; Franusich, Michael D.

2010-10-12

An online data driven computerized procedures system that guides an operator through a complex process facility's operating procedures. The system monitors plant data, processes the data and then, based upon this processing, presents the status of the current procedure step and/or substep to the operator. The system supports multiple users and a single procedure definition supports several interface formats that can be tailored to the individual user. Layered security controls access privileges and revisions are version controlled. The procedures run on a server that is platform independent of the user workstations that the server interfaces with and the user interface supports diverse procedural views.

Development of a Desktop Simulator for APR1400 Nuclear Power Plant

International Nuclear Information System (INIS)

Lee, J. B.

2016-01-01

It is essential for utilities to possess a full-scope simulator for operator training and operation test for operators. But it is very expensive and sometimes lack of fidelity if processes of developing the simulator and designing the plant are in parallel. It is due to the situation that simulator development stage sometimes precedes the plant design stage and modifications may occur to the design of the plant in construction stage. In an attempt to build a low cost and efficient simulator, a desktop simulator has been developed. This model is described herein. Using desktop simulators for training operators is an efficient method for familiarizing operators with their plant’s operation. A low cost and efficient desktop simulator for APR1400 has been developed, and brief features are introduced here. It is configured to mimic a full-scale simulator, and can be used for operators to be familiarized to their plant’s operation. Since the size of the simulator is small enough to be fit in a desk, it can be used in a classroom or in an office at any time. It can also be used to evaluate design changes or modifications of the plant before implementing them to the plant

Development of a Desktop Simulator for APR1400 Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Lee, J. B. [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

It is essential for utilities to possess a full-scope simulator for operator training and operation test for operators. But it is very expensive and sometimes lack of fidelity if processes of developing the simulator and designing the plant are in parallel. It is due to the situation that simulator development stage sometimes precedes the plant design stage and modifications may occur to the design of the plant in construction stage. In an attempt to build a low cost and efficient simulator, a desktop simulator has been developed. This model is described herein. Using desktop simulators for training operators is an efficient method for familiarizing operators with their plant’s operation. A low cost and efficient desktop simulator for APR1400 has been developed, and brief features are introduced here. It is configured to mimic a full-scale simulator, and can be used for operators to be familiarized to their plant’s operation. Since the size of the simulator is small enough to be fit in a desk, it can be used in a classroom or in an office at any time. It can also be used to evaluate design changes or modifications of the plant before implementing them to the plant.

Improved Design of PECS to reduce Flow Instability for EU-APR1400

International Nuclear Information System (INIS)

Hwang, Do Hyun; Lee, Keunsung

2014-01-01

For EU-APR1400, PECS (Passive Ex-vessel corium retaining and Cooling System), so-called core catcher, was adopted to keep the integrity of basemat in containment by preventing MCCI (Molten Core Concrete Interaction) through retaining core debris and cooling corium outside the reactor vessel. In this paper, the improved design of PECS is presented to increase coolability by reducing flow instability in the region of cooling channel. In this paper, flow instability analysis was carried out using CFD code to find out the most improved design of PECS, which is to increase coolability by reducing bubble entrainment in the region of cooling channel. The reduction of bubble entrainment in the downcomer facilitates higher mass flow rates in the downcomer. Among presented four designed for the downcomer of PECS, the superstep design shows the highest mass flow rate and the lowest gas holdup in the downcomer as well as in the cooling channel. Compared with the existing design, the elimination of the horizontal part and the addition of an extra space above the vertical entrance to the downcomer seem to help the separation of the vapor

Fuel assembly design for APR1400 with low CBC

Energy Technology Data Exchange (ETDEWEB)

Hah, Chang Joo, E-mail: changhah@kings.ac.kr [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-04-29

APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

Analysis for the Effects of Grid Voltage Degradation on APR1400 Operation, Case Study for Egypt

International Nuclear Information System (INIS)

Hassan, Mostafa Ahmed Fouad; Koo, Chang Choong

2015-01-01

Egypt is one of the countries planning to introduce a NPP into its electrical power system. Although the Egyptian power system has sufficient capacity to integrate any commercially available nuclear unit as the total installed capacity of the power system is more than 32GWe, which is more than 10 times capacity of any nuclear unit in the range of 1000 to 1700MWe, the system is vulnerable to extreme voltage variations, especially voltage degradation during peak load conditions. These conditions can lead to voltage collapse if a counter measure, usually load shedding, is not taken in a proper time. Hence, it is necessary to analyze the effect of such conditions on the safe and economic operation of the NPP. In this paper we analyzed the effects of grid voltage degradation on the safe and economic operation of the Advanced Power Reactor (APR1400) to determine any adverse effects on the plant auxiliary loads while operating in the Egyptian power system. In this paper the effects of grid voltage degradation on the safe and economic operation of APR1400 were investigated taking into account, generator operating limits, plant safety requirements, operation modes and loading categories in order to determine any adverse effect on the plant auxiliary loads while operating in the Egyptian power system. The results of the load flow and motor starting analysis demonstrated that during normal operation the automatic voltage regulator and transformers OLTCs can mitigate the effect of grid voltage degradation without any detrimental effect on the plant auxiliary loads. During the highly unlikely LOCA condition if the grid voltage degraded below 95%, the degraded voltage relays at Class 1E 4.16 kV buses will trip the supply and load breakers and reconnect the required safety loads to the EDG after 4 minutes time delay. During this period the safety loads required for LOCA can be started and accelerated to their rated speed safely even in the worst case of expected degraded voltage

Mobile Computerized Procedure System for the Improved Situation Awareness among Field Workers and Main Control Room Operators

International Nuclear Information System (INIS)

Lee, Sungjin; Seong, Nokyu; Jung, Yeonsub

2013-01-01

This paper explains the technical issues of the mobile CPS (m-CPS) in the conceptual development stage. The proposed m-CPS can be developed with many recent technologies in the various industry fields based on CRI CPS. Performance and stability of CRI CPS have been tested several times until now. It is expected that more human errors can be reduced by the m-CPS through improved situation awareness and human performance tools for local workers and MCR operators. Some considerations as well as mentioned ones will be reflected in the m-CPS development. Human errors can occur during the test and maintenance of the generator, safety injection system and reactor coolant inventory in nuclear power plants (NPPs). Most of human errors have been occurred by the omission of the prevention techniques such as the self-check, the peer-check, the concurrent verification and etc. Another important reason is the insufficient information sharing among main control room (MCR) operators and field workers. Various field service automation tools have been developed with recent information technology in many countries. APR1400 computerized procedure system (CPS) has been developed for the MCR operators of Shin-Kori 3 and 4 units. Especially, the concurrent verification support design is applied in the construction project of Shin-Hanul 1 and 2 CPS. KHNP central research institute (CRI) expects that the extended application of CPS including the field activity, that is a kind of mobile CPS, can enhance the reduction of human errors

Evaluation on thermal-hydraulic characteristics for passive safety device of APR1400

Energy Technology Data Exchange (ETDEWEB)

Yoo, Seong Yeon; Lee, S. H.; Son, M. K. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of); Jee, M. S.; Chung, M. H. [Chungnam National Univ., Taejon (Korea, Republic of)

2001-07-15

To establish evaluation and verification guideline for the APR1400, thermal-hydraulic characteristics for fuel rod bundle, reactor vessel and fluidic device is analyzed using FLUENT. Scope and major results of research are as follows : Thermal-hydraulic characteristics for nuclear fuel rod bundle: design data for nuclear fuel rod bundle and structure are surveyed, and 3 x 3 sub-channel model is adopted to investigate the fluid flow and heat transfer characteristics in fuel rod bundle. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions. Thermal-hydraulic characteristics for reactor vessel: reactor vessel design data are surveyed to develop numerical model. Porous media model is applied for fuel rod bundle, and full-scale, three dimensional simulation is performed at actual operating conditions. Distributions of velocity, pressure and temperature are discussed. Flow characteristics for fluidic device: three dimensional numerical model for fluidic device is developed, and numerical results are compared with experimental data obtained at KAERI in order to verify numerical simulation. In addition, variation of flow rate is investigated at various elapsed times after valve operating, and flow characteristics is analyzed at low and high flow rate conditions, respectively.

Modeling of melt retention in EU-APR1400 ex-vessel core catcher

Energy Technology Data Exchange (ETDEWEB)

Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

2012-07-01

A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

A Study on the Coupled FEM-Analysis for Reactor Vessel Lower Head of APR1400 under the Severe Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyonam; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

For the stabilization of the RPV the in-vessel retention strategy with external reactor vessel cooling (IVR-ERVC) is adopted in APR1400. Under this severe accident condition, a good understanding of the mechanical behavior of the reactor vessel lower head (RVLH) is necessary both for verification of structural integrity and for improving the design applying appropriate accident mitigation strategies. The purpose of this study is to develop the analysis method of the RVLH with thermo-mechanical analysis using FEM tool (ANSYS v.15) in case of core-melting severe accident condition, and then analyze the RVLH of APR1400 including creep behavior. The plastic strain can be the major cause of lower head failure on the reactor vessel, and the creep cannot be not negligible factor of the failure under the severe accident condition. In the study, we applied constant convection coefficient at assumed temperature on the outside wall of RPV and substitute creep data of SA-508. In addition, it was found that the steel ablation at the interface between corium and vessel steel is not only a thermal phenomenon in the METCOR experiments. Corrosion processes and the formation of eutectics lead to the erosion of the vessel steel at temperatures that are significantly lower than the melting temperature of steel. It called thermo-chemical attack of the corium (corrosion). Reduced wall thickness because of the thermo-chemical effect by corium increase the equivalent plastic strain, and decrease the minimum time to reach 20% creep strain.

Design of A Vibration and Stress Measurement System for an Advanced Power Reactor 1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program

International Nuclear Information System (INIS)

Ko, Doyoung; Kim, Kyuhyung

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

Directory of Open Access Journals (Sweden)

DO-YOUNG KO

2013-04-01

Full Text Available In accordance with the US Nuclear Regulatory Commission (US NRC, Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP has been developed for an Advanced Power Reactor 1400 (APR1400. The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment. Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

The Development and Evaluation of Inherent RPCS for the APR1400

Energy Technology Data Exchange (ETDEWEB)

Choi, Jae D.; Ryu, Seok H.; Jung, Dong C.; Kim, Joon S.; Baek, Byung C.; Sung, Song K.; You, Guk J. [KNF, Daejeon (Korea, Republic of); Kim, Han G. [KHNP, Daejeon (Korea, Republic of); Chi, Sung G. [KOPEC, Daejeon (Korea, Republic of)

2008-10-15

The APR1400 RPCS (Reactor Power Cutback System) is designed to rapidly reduce the core power to eliminate the need for a reactor trip following a large load rejection or a loss of two main feedwater pumps at high power. GDC (General Design Criteria) 25 says 'Protection system requirements for reactivity control malfunctions. The protection system shall be designed to assure that SAFDL (Specified Acceptable Fuel Design Limits) are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.' In order to comply GDC 25, CPCS (Core Protection Calculator System) will apply a big power penalty({approx} 1.3) to determine the minimum DNBR (Departure from Nucleate Boiling Ratio), the maximum LPD(Local Power Density) and reactor may immediately trip by CPCS low DNBR and high LPD during high power operation for 12-finger CEA drop. The purpose of this study is to develop the inherent RPCS to avoid unwanted reactor trips due to a 12-finger CEA drop. In order to accomplish this purpose, the CPCS should be modified to send RPCS actuation signal and not to apply power penalty due to 12-finger CEA drop for a shot period. During this period which is determined by assessment of safety, the SAFDL will not be violated without any CPCS trip function. The system and CPCS performance is evaluated to verify that CPCS trip does not occurred during 12-finger CEA drop event.

Study on computerized presentation of emergency procedures of a nuclear plant (step 2). The guidelines of the prototype of the computerized procedure presentation system

International Nuclear Information System (INIS)

Niwa, Yuji; Hollnagel, E.; Iwaki, Toshio.

1995-01-01

New methods of information presentation and interface design are changing the working conditions in the modern Nuclear Power Plant (NPP) control room. One area receiving considerable attention is that of Emergency Operating Procedures (EOPs), which plays an essential role in NPPs. Conventionally procedures are presented in a hard copy form. However developments in information technology have offered new opportunities for the computerization of such procedures. Consideration for the first stage of computerization should be focused upon the presentation of procedures. The specification of the computerized presentation of procedures is discussed with respect to the issues which were central to the project: navigation through procedures; formatting and presentation of procedures; and process monitoring. Issues that would be included in more advanced systems, such as help and explanation facilities features, and process linking, are also discussed. This paper deals with the specific design guidelines that were implemented for the computerization of procedure presentation. Issues of principal concern that were identified from this experience are highlighted, such as the relationship between procedure presentation and format; the registration of progress through a procedure; compensation for the limitations of computer displays versus printed documents; and the way in which the added capabilities of computerized presentation can be generally utilized in the operators' working environment. (author)

Analysis of Hydrogen Concentration Distribution during an SBO Accident for Shin-Ulchin APR1400

Energy Technology Data Exchange (ETDEWEB)

Kim, Jongtae; Hong, Seong Wan [Korea Atomic energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

To prohibit the accumulation of hydrogen, the containment volume is considered to reduce the hydrogen concentration, or hydrogen mitigation devices such as PARs or igniters are installed in the containment. In the case of the Fukushima NPPs, the applied strategy for the hydrogen safety is the use of a containment venting system (CVS). In this way, the hydrogen accumulated in the containment vessel is vented into the environment. One of the causes of the hydrogen explosions occurring in the containment buildings of the Fukushima NPPs is expected to be the failure of the venting system. The hydrogen was therefore easily accumulated in the containment building. It is uncertain what the ignition source for the hydrogen combustion was during the accident. However, it is not too conservative to assume that an ignition source exists at any time and any place in a containment during a core-melt accident. Shin-Ulchin 1 and 2, which are construction plants of an APR 1400, are two of the newest NPPs in Korea. They have many features to enhance the safety margin during a design-based and beyond-design-based accident. One of them is the in-containment refueling water storage tank (IRWST) located inside the containment. It is used as a sink/source for feed-bleed operation. When the core is damaged along an accident progression, the hydrogen generated in the RPV can be released into the IRWST of the APR1400 with steam and water. From a previous study, it was found that a highly concentrated hydrogen/air mixture can be developed if the hydrogen is released into the IRWST. In the case of Shin-Ulchin 1 and 2, the hydrogen mitigation strategy during a high-pressure accident such as a station blackout (SBO) is changed by installing a 3-way valve. When a severe accident management (SAM) for the plant is initiated, the flow path from a pressurizer to the IRWST is changed into a steam-generator (S/G) compartment by turning the 3-wat valve actively (pilot operated). By doing so, it is

Optimization Strategy of the APR+ BOP Technical Specifications

International Nuclear Information System (INIS)

Cho, Yoon Sang; Lee, Jae Gon; Han, Sung Heum

2016-01-01

The BOP is one of the key factors for successful project implementation of NPP. In constructing the APR1400 NPP, the BOP procurement has been one of the biggest concerns. Due to the design changes and increased capacity of equipment in NPP, lots of BOPs should be ‘first supplied equipment’ [hereinafter, ‘FSE’]. The manufacture-ability, and the performances of FSEs have not been fully proved and tested, manufacturers and suppliers are requested to submit Reports for Equipment Qualification Evaluation in accordance with 10 CFR 50.49, IEEE 323. They need at least 1-2 years’ tests for Environment Qualification (EQ) and Seismic Qualification (SQ). This study is focused how to prepare the BOP purchase specifications in order to control the FSEs, especially in safety class equipment. With the optimization plan for BOP packages of this study, the FSEs’ occurrence can be reasonably controlled as low as possible. For successful NPP project, the concerns in procuring BOPs shall be fully analyzed beforehand. Now Korea is preparing new era of APR+, with closing the time of APR1400. The technical specification of APR+ BOPs can be developed and prepared successfully and very effectively according to this optimization plan. This will be a great contribution not only in constructing APR+ in time, but also in exporting APR+ overseas, all over the world in the future

Optimization Strategy of the APR+ BOP Technical Specifications

Energy Technology Data Exchange (ETDEWEB)

Cho, Yoon Sang; Lee, Jae Gon; Han, Sung Heum [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The BOP is one of the key factors for successful project implementation of NPP. In constructing the APR1400 NPP, the BOP procurement has been one of the biggest concerns. Due to the design changes and increased capacity of equipment in NPP, lots of BOPs should be ‘first supplied equipment’ [hereinafter, ‘FSE’]. The manufacture-ability, and the performances of FSEs have not been fully proved and tested, manufacturers and suppliers are requested to submit Reports for Equipment Qualification Evaluation in accordance with 10 CFR 50.49, IEEE 323. They need at least 1-2 years’ tests for Environment Qualification (EQ) and Seismic Qualification (SQ). This study is focused how to prepare the BOP purchase specifications in order to control the FSEs, especially in safety class equipment. With the optimization plan for BOP packages of this study, the FSEs’ occurrence can be reasonably controlled as low as possible. For successful NPP project, the concerns in procuring BOPs shall be fully analyzed beforehand. Now Korea is preparing new era of APR+, with closing the time of APR1400. The technical specification of APR+ BOPs can be developed and prepared successfully and very effectively according to this optimization plan. This will be a great contribution not only in constructing APR+ in time, but also in exporting APR+ overseas, all over the world in the future.

Drop size measurements and entrainment in APR1400 during LBLOCA reflood phase

Energy Technology Data Exchange (ETDEWEB)

Lee, Eo Hwak

2010-02-15

A study has been performed to investigate droplet size in the nuclear reactor of APR1400 during LBLOCA reflood phase and to develop droplet entrainment and deposition models for SPACE (Safety and Performance CodE) which is a safety analysis tool for PWR being developed in Korea. A freezing technique for measuring the size of droplets was developed to obtain the droplet size distribution in horizontal annular flow in a pipe with a 37.1 mm diameter. Droplets are frozen by using an extremely low temperature nitrogen gas with liquid film extraction. They are then photographed with a microscope and a CCD camera and measured by means of an image process. The results are compared with various experimental data. The droplet sizes measured by the freezing technique are comparable with those measured by other methods at a high superficial air velocity (of 50 m/s). However, because of the film extraction problem, the droplet sizes measured at a low superficial air velocity of less than 40 m/s are higher than those measured by other methods. A present method suggested for predicting the Sauter mean diameter is based on the maximum droplet size correlation for the experimental data, with and without liquid film extraction. The average droplet size is remarkably smaller downstream of the liquid film extractor because large droplets from the liquid film are excluded. In order to understand and to predict a heat transfer between superheated steam and droplets properly during reflood phase of LBLOCA, it is very important to measure broken droplet sizes by spacer grids. A study, therefore, has been performed to investigate droplet size in rod bundles with spacer grids and to develop a spacer grid droplet breakup model for safety analysis codes. Experiments were conducted with liquid droplets (SMD of 300∼700 μm) and various spacer grids at superficial air velocity of 10 m/s and 20 m/s based on FLECHT SEASET. The test channel and the grids were heated to 150 .deg. C to prevent

Drop size measurements and entrainment in APR1400 during LBLOCA reflood phase

International Nuclear Information System (INIS)

Lee, Eo Hwak

2010-02-01

A study has been performed to investigate droplet size in the nuclear reactor of APR1400 during LBLOCA reflood phase and to develop droplet entrainment and deposition models for SPACE (Safety and Performance CodE) which is a safety analysis tool for PWR being developed in Korea. A freezing technique for measuring the size of droplets was developed to obtain the droplet size distribution in horizontal annular flow in a pipe with a 37.1 mm diameter. Droplets are frozen by using an extremely low temperature nitrogen gas with liquid film extraction. They are then photographed with a microscope and a CCD camera and measured by means of an image process. The results are compared with various experimental data. The droplet sizes measured by the freezing technique are comparable with those measured by other methods at a high superficial air velocity (of 50 m/s). However, because of the film extraction problem, the droplet sizes measured at a low superficial air velocity of less than 40 m/s are higher than those measured by other methods. A present method suggested for predicting the Sauter mean diameter is based on the maximum droplet size correlation for the experimental data, with and without liquid film extraction. The average droplet size is remarkably smaller downstream of the liquid film extractor because large droplets from the liquid film are excluded. In order to understand and to predict a heat transfer between superheated steam and droplets properly during reflood phase of LBLOCA, it is very important to measure broken droplet sizes by spacer grids. A study, therefore, has been performed to investigate droplet size in rod bundles with spacer grids and to develop a spacer grid droplet breakup model for safety analysis codes. Experiments were conducted with liquid droplets (SMD of 300∼700 μm) and various spacer grids at superficial air velocity of 10 m/s and 20 m/s based on FLECHT SEASET. The test channel and the grids were heated to 150 .deg. C to prevent

Two-phase flow parameters of a downcomer boiling during a postulated reflood phase of APR1400; Parametres d'ecoulement diphasique dans un ebullition a la cuve en acier pendant un refroidissement d'un APRP de APR1400

Energy Technology Data Exchange (ETDEWEB)

Byong-Jo, Yun; Dong-Jin, Euh; Chul-Hwa, Song [Korea Atomic Energy Research Inst. (Korea, Republic of)

2009-07-01

Downcomer boiling phenomena is one of the key issues for a postulated large-break LOCA (LBLOCA) in a conventional pressurized water reactor, because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region of the APR1400, a test program for a downcomer boiling is being progressed for the reflood phase of a postulated LBLOCA. Test was performed in a one side heated rectangular test channel which was designed by adopting a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length was reduced 47.08-fold. The test consists of two steps: (I) for the global two-phase flow parameters and (II) for the local two-phase flow parameters. The step-I test has already been completed. In the present paper, the experimental results of the step-II test are introduced. (authors)

Consolidation and Decomposition of APR1400 NRC Design Certification Processes for Collaborative and Accelerated Processing

Energy Technology Data Exchange (ETDEWEB)

Cho, Young Chul [FNC Technology Co., Yongin (Korea, Republic of); Kang, Deog Ji [KNHP, Daejeon (Korea, Republic of)

2016-05-15

KEPCO and KHNP are conducting APR1400 Design Certification from NRC. For the proper management of processes and information, a system called RIS (Regulatory Information management System) has been implemented by FNC from 2014, and it is on the final stage. In retaining the certification from NRC, RIS will be a very essential role by providing platform for collaborative and accelerated processing of responses to RAI (Request of Additional Information). Preparation of responses to RAI with this kind of systematic approach may be the first in the world. Westinghouse is doing manually and using MS Excel to collect the processing status. Where as, RIS will enable each member can do his own job and collect the status automatically. In this paper, how collaborative and accelerated processing of responses to RAI can be enabled will be described, and further enhancements will also be discussed. It handles MS Word directory with the help of VSTO. And with the help of Aspose.Total for .NET, the prepared response to RAI meets NRC's requirements. Through some further work and direct integration with requirement management solution, RIS can be expanded to cover a prior impact notice in case of DCD being altered.

Preliminary COM3D Analysis for H{sub 2} Combustion in the APR1400 Containment

Energy Technology Data Exchange (ETDEWEB)

Kang, Hyung Seok; Kim, Jongtae; Kim, Sang-Baik; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In order to assure the containment integrity of APR1400, it is necessary to evaluate an overpressure buildup resulting from a propagation of hydrogen flame along the structure and wall in the containment during a severe accident using the COM3D. Korea Atomic Energy Research Institute (KAERI) established a numerical analysis system for evaluating an overpressure buildup owing to a hydrogen combustion in the containment of a nuclear power plant by importing COM3D from KIT in Germany. The COM3D has been currently used in the simulations of the combustions and explosions, together with the system of industrial risk mitigation of hydrogen and burnable gases in nuclear containment and auxiliary buildings. KAERI performed the hydrogen combustion analysis using the COM3D code with the initial hydrogen distribution calculated by the GASFLOW under assumption of 85% metal-water reaction in the reactor vessel. From the COM3D results, we can know that the pressure buildup was about 100 kPa because the flame speed was not increased above 1000 m/s and the pressure wave passed through the open spaces in the large containment.

Current Human Reliability Analysis Methods Applied to Computerized Procedures

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring

2012-06-01

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room (Fink et al., 2009). Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of enhanced ease of use and easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

APR1400 Containment Simulation with CONTAIN code

Energy Technology Data Exchange (ETDEWEB)

Hwang, Moon Kyu; Chung, Bub Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-05-15

The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

APR1400 Containment Simulation with CONTAIN code

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Chung, Bub Dong

2010-01-01

The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

Design Characteristics of Soft Control for APR+

Energy Technology Data Exchange (ETDEWEB)

Kim, Yongsoo; Sung, Chanho; Jung, Yeonsub [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In the global nuclear industry, the design requirements have been largely divided into the US and the European ones. The APR+ design, based on the proven APR1400 design, was developed to fully comply with the US regulatory and utility requirements. It was also developed to improve the technical competitiveness and enlarge the export market share through design uniqueness and the improvement of safety. The robust design of safety and its validation were requested Post-Fukushima accident. So MMIS design and validation for reinforcing safety were accomplished in APR+ development project. In this paper, the design characteristics of soft control for APR+ are presented to enhance the safety of MMIS design. In this paper, the major design characteristics of soft control for APR+ were introduced to increase the safety of MMIS design. In addition to the design characteristics of soft control described above, the following should be investigated closely for APR+ to get into the global nuclear industry.

Design Characteristics of Soft Control for APR+

International Nuclear Information System (INIS)

Kim, Yongsoo; Sung, Chanho; Jung, Yeonsub

2016-01-01

In the global nuclear industry, the design requirements have been largely divided into the US and the European ones. The APR+ design, based on the proven APR1400 design, was developed to fully comply with the US regulatory and utility requirements. It was also developed to improve the technical competitiveness and enlarge the export market share through design uniqueness and the improvement of safety. The robust design of safety and its validation were requested Post-Fukushima accident. So MMIS design and validation for reinforcing safety were accomplished in APR+ development project. In this paper, the design characteristics of soft control for APR+ are presented to enhance the safety of MMIS design. In this paper, the major design characteristics of soft control for APR+ were introduced to increase the safety of MMIS design. In addition to the design characteristics of soft control described above, the following should be investigated closely for APR+ to get into the global nuclear industry

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

OpenAIRE

KO, DO-YOUNG; KIM, KYU-HYUNG

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

Quantitative Safety Impact of Severe Accident Management Systems for EU-APR during Low Power Shutdown Operation

International Nuclear Information System (INIS)

Lee, Keunsung; Hwang, Do Hun; Chang, Hyun-bin

2016-01-01

In order to enlarge and to diversify the export market of APR1400, the EU-APR design was developed based on the APR1400 design to comply with the latest version of the European Utility Requirements (EUR) revision D. The EU-APR design has the distinguished and advanced severe accident management systems taken from the APR1400 to obtain a containment integrity for the beyond design basis accident, such as the Passive Ex-vessel retaining and Cooling System (PECS), the Severe Accident Containment Spray System (SACSS) and the Containment Filtered Vent System (CFVS). The risk associated with the nuclear power plant can be identified through the Probabilistic Safety Assessment (PSA). In the EUR chapter 1 and 17 of volume 2, the Criteria for Limited Impact (CLI) should be applied to the Level 2 PSA as a risk metrics. The fraction of exceeding CLI for the EU-APR during LPSD operation was calculated as 4.52% of the CDF under the condition that all severe accident management systems are credited. The PECS, SACSS and CFVS are considered as the severe accident management system which is EU-APR dedicated system. The exemption of each system leads to increase the fraction of exceeding CLI to 54.18%, 89.74% and 21.32% respectively. In case if all these systems are unavailable, the fraction of exceeding CLI is increased to 100%. The most effective system is the SACSS that the system reduces containment pressure and temperature

Quantitative Safety Impact of Severe Accident Management Systems for EU-APR during Low Power Shutdown Operation

Energy Technology Data Exchange (ETDEWEB)

Lee, Keunsung; Hwang, Do Hun [KHNP CRI, Daejeon (Korea, Republic of); Chang, Hyun-bin [Future and Challenge Technology Co., Yongin (Korea, Republic of)

2016-10-15

In order to enlarge and to diversify the export market of APR1400, the EU-APR design was developed based on the APR1400 design to comply with the latest version of the European Utility Requirements (EUR) revision D. The EU-APR design has the distinguished and advanced severe accident management systems taken from the APR1400 to obtain a containment integrity for the beyond design basis accident, such as the Passive Ex-vessel retaining and Cooling System (PECS), the Severe Accident Containment Spray System (SACSS) and the Containment Filtered Vent System (CFVS). The risk associated with the nuclear power plant can be identified through the Probabilistic Safety Assessment (PSA). In the EUR chapter 1 and 17 of volume 2, the Criteria for Limited Impact (CLI) should be applied to the Level 2 PSA as a risk metrics. The fraction of exceeding CLI for the EU-APR during LPSD operation was calculated as 4.52% of the CDF under the condition that all severe accident management systems are credited. The PECS, SACSS and CFVS are considered as the severe accident management system which is EU-APR dedicated system. The exemption of each system leads to increase the fraction of exceeding CLI to 54.18%, 89.74% and 21.32% respectively. In case if all these systems are unavailable, the fraction of exceeding CLI is increased to 100%. The most effective system is the SACSS that the system reduces containment pressure and temperature.

A computerized system for improved management and execution of plant procedures

International Nuclear Information System (INIS)

Lipner, M.H.; Orendi, R.G.; Impink, A.J.; Mundy, R.A.

1987-01-01

The authors have shown how the deficiencies reported in this paper can be resolved by using the Computerized Procedures System. The Westinghouse Computerized Procedures System solves many of the human performance problems associated with procedures. It has been designed to assist plant operators in executing procedures more efficiently and cost effectively. The system brings all of the necessary information to one location for easy and continuous assessment of plant conditions and permits the operator to concentrate on the big picture

Development of the APR+ Auxiliary Building General Arrangement (GA)

International Nuclear Information System (INIS)

Moon, Hyung Keun; Park, Young Sheop; Kang, Yong Chul

2011-01-01

The general arrangement (GA) drawing of a nuclear power plant is the most basic drawing which contains all of the plant equipment, systems, and rooms. Therefore, it should be issued at an early design stage to provide the contours of the overall plant structure. This type of drawing is typically used widely throughout the design stages. The development project of APR+ (Advanced Power Reactor+), as a succeeding model of the APR1400 (Advanced Power Reactor 1400) design, has its own GA that encompasses all of its power buildings. This was developed starting in October of 2009. Among several of the buildings in this design, the Auxiliary Building (AB) is one of the most important buildings to produce electricity, and to protect against undesirable radiation emissions. This paper focuses on the design characteristics of the general arrangement of the AB

Experimental studies of computerized procedures and team size in nuclear power plant operations

International Nuclear Information System (INIS)

Huang, F.-H.; Hwang, S.-L.

2009-01-01

The operation of a nuclear power plant is so complex that it requires teamwork. To support team performance, a system need to provide all team members integrated information displays as well as decision aids (e.g., computerized procedures). Two experiments were conducted to investigate the effects of computerized procedures and team size on operating performance. Forty-five participants were involved in the experiments. Each participant executed decision and action tasks to deal with alarm signals, while detecting occasional system errors in the interface. Results showed that effects of computerized procedures were significant on various performance indicators, such as operation time, operation errors, and learning effect, and that two operators would be a satisfactory size in the teamwork system providing computerized procedures

CONSIDERATIONS FOR THE TREATMENT OF COMPUTERIZED PROCEDURES IN HUMAN RELIABILITY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; David I. Gertman

2012-07-01

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room. Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Development of Alarm System link Drawing for Operation Support for APR1400 Digital Main Control Room

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki-Hwan [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Digitalized MMI(Man-Machine Interface) including Digital Main Control Room(MCR) and digital I and C system was being applied for SKN 3 and 4 Nuclear Power Plant(NPP) and subsequent APR1400 NPP type. But, operators can not easily find instrument for alarm immediately. Therefore, Alarm system is required to easily find instrument for Alarm. For this implementation, we will plan system design considering design feature without affecting network load and CPU load. We have developed Alarm system link drawing for digital MCR. Operators of the digitalized MCR navigates from their consoles to the drawings related to the plant alarms and their instruments or the operation status. Such method gives cognitive load to the operators having to travel to different locations in finding the related information. Screen Sharing System, which is the fundamental technique for Drawing Interconnection Alarm System is close to completion, and it should be functionally tested and verified by the human factor engineering. For the actual application to the operating plants, the drawings to be interconnected to the alarms and the opinions from the operators/maintenance departments for designating alarm number should be surveyed, Also, another function that allows the access to the alarm related drawings not only from the MCR but also from the other offices.

Development of Alarm System link Drawing for Operation Support for APR1400 Digital Main Control Room

International Nuclear Information System (INIS)

Kim, Ki-Hwan

2016-01-01

Digitalized MMI(Man-Machine Interface) including Digital Main Control Room(MCR) and digital I and C system was being applied for SKN 3 and 4 Nuclear Power Plant(NPP) and subsequent APR1400 NPP type. But, operators can not easily find instrument for alarm immediately. Therefore, Alarm system is required to easily find instrument for Alarm. For this implementation, we will plan system design considering design feature without affecting network load and CPU load. We have developed Alarm system link drawing for digital MCR. Operators of the digitalized MCR navigates from their consoles to the drawings related to the plant alarms and their instruments or the operation status. Such method gives cognitive load to the operators having to travel to different locations in finding the related information. Screen Sharing System, which is the fundamental technique for Drawing Interconnection Alarm System is close to completion, and it should be functionally tested and verified by the human factor engineering. For the actual application to the operating plants, the drawings to be interconnected to the alarms and the opinions from the operators/maintenance departments for designating alarm number should be surveyed, Also, another function that allows the access to the alarm related drawings not only from the MCR but also from the other offices

Onset of liquid droplet entrainment on a direct vessel injection system for APR1400

Energy Technology Data Exchange (ETDEWEB)

Kim, Han-sol; Lee, Jae-Young [Handong Global University, Pohang (Korea, Republic of); Kim, Jong-Rok; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this research, a series of visualization works was conducted to understand droplet entrainment of the flow pattern generated in direct vessel injection system(DVI) of Korea nuclear power plant, APR 1400. In the emergency situation of a nuclear power plant, reliability of DVI cooling can be an important issue. It is known that, the amount and the rate of entrainment during the DVI cooling process can significantly affect the total heat removal. To visualize the film Reynolds number closely related with onset of droplet entrainment induced by falling film flow and lateral air flow in a small gap, confocal chromatic sensing method for measuring accurately film thickness and depth averaging particle image velocimetry for film velocity were used. The results have been post processed 4G Insight software. By measuring two dimensional film Reynolds number, we can predict the onset of droplet entrainment and obtain visible breakup region intuitively. To visualize the droplet entrainment induced by falling film flow and lateral air flow in a small gap, shadowgraph method with CCD camera (2200fps, 1280 pixel X 800 pixel, ) on coated plate with super water-repellent agent was used. The results have been post processed using 4G Insight software. By measuring two dimensional film Reynolds number, we can predict the onset of droplet entrainment and obtain visible breakup region intuitively. By adopting both super hydrophobic coating method and shadowgraph method, entrainment in a narrow gap was successfully visualized that has rarely performed before and meaningful results for DVI system research fields have been made.

Onset of liquid droplet entrainment on a direct vessel injection system for APR1400

International Nuclear Information System (INIS)

Kim, Han-sol; Lee, Jae-Young; Kim, Jong-Rok; Euh, Dong-Jin

2016-01-01

In this research, a series of visualization works was conducted to understand droplet entrainment of the flow pattern generated in direct vessel injection system(DVI) of Korea nuclear power plant, APR 1400. In the emergency situation of a nuclear power plant, reliability of DVI cooling can be an important issue. It is known that, the amount and the rate of entrainment during the DVI cooling process can significantly affect the total heat removal. To visualize the film Reynolds number closely related with onset of droplet entrainment induced by falling film flow and lateral air flow in a small gap, confocal chromatic sensing method for measuring accurately film thickness and depth averaging particle image velocimetry for film velocity were used. The results have been post processed 4G Insight software. By measuring two dimensional film Reynolds number, we can predict the onset of droplet entrainment and obtain visible breakup region intuitively. To visualize the droplet entrainment induced by falling film flow and lateral air flow in a small gap, shadowgraph method with CCD camera (2200fps, 1280 pixel X 800 pixel, ) on coated plate with super water-repellent agent was used. The results have been post processed using 4G Insight software. By measuring two dimensional film Reynolds number, we can predict the onset of droplet entrainment and obtain visible breakup region intuitively. By adopting both super hydrophobic coating method and shadowgraph method, entrainment in a narrow gap was successfully visualized that has rarely performed before and meaningful results for DVI system research fields have been made

Evaluation of the computerized procedures Manual II (COPMA II)

International Nuclear Information System (INIS)

Converse, S.A.

1995-11-01

The purpose of this study was to evaluate the effects of a computerized procedure system, the Computerized Procedure Manual II (COPMA-II), on the performance and mental workload of licensed reactor operators. To evaluate COPMA-II, eight teams of two operators were trained to operate a scaled pressurized water reactor facility (SPWRF) with traditional paper procedures and with COPMA-II. Following training, each team operated the SPWRF under normal operating conditions with both paper procedures and COPMA-II. The teams then performed one of two accident scenarios with paper procedures, but performed the remaining accident scenario with COPMA-II. Performance measures and subjective estimates of mental workload were recorded for each performance trial. The most important finding of the study was that the operators committed only half as many errors during the accident scenarios with COPMA-II as they committed with paper procedures. However, time to initiate a procedure was fastest for paper procedures for accident scenario trials. For performance under normal operating conditions, there was no difference in time to initiate or to complete a procedure, or in the number of errors committed with paper procedures and with COPMA-II. There were no consistent differences in the mental workload ratings operators recorded for trials with paper procedures and COPMA-II

Development of the APR1400 model for countercurrent natural circulation in hot leg and steam generator under station blackout

International Nuclear Information System (INIS)

Park, Sang Gil; Kim, Han Chul

2012-01-01

In order to analyze severe accident phenomena, Korea Institute of Nuclear Safety (KINS) made a MELCOR model for APR1400 to examine natural circulation and creep rupture failure in the Reactor Coolant System (RCS) under station blackout (SBO). In this study, we are trying to advance the former model to describe natural circulation more accurately. After Fukushima accident, the concerns of severe accident management, assuring the heat removal capability, has risen for the case when the SBO is happened and there are no more electric powers to cool down decay heat. Under SBO there are three kinds of natural circulation which can delay the core heatup. One is in vessel natural circulation in the upper plenum of reactor vessel and the second is countercurrent natural circulation in hot leg through steam generator tubes and the last is full loop natural circulation when the reactor coolant pump loop seal is cleared and reactor coolant pump sealing is damaged by high temperature and high pressure. Among them this study focuses on the countercurrent natural circulation model using MELCOR1.8.6

Air clearing pressure oscillation produced in a quenching tank by a prototype unit cell sparger of the APR1400

International Nuclear Information System (INIS)

Cho, Seok; Song, Chul-Hwa; Park, Choon-Kyong; Kim, Hwan-Yeol; Baek, Won-Pil

2008-01-01

KAERI has performed a series of experiments to investigate the performance of a prototype sparger for the APR1400 in view of a dynamic load oscillation with a variation of the test conditions such as a discharged air mass, a submergence of the sparger, the valve opening time, and the pool temperature during an air clearing phase. The air mass and pool temperature are in the range of 0.8-1.5 kg and 20-90 o C, respectively. The valve opening time can be adjusted within the range of 0.6-1.8 s. The maximum positive pressure amplitude, which is observed at the bottom of the quenching tank, is increased with the maximum header pressure of the sparger. The valve opening time has a considerable effect on the maximum amplitude. As the opening time decreases, the maximum amplitude at the tank wall is increased. Air mass and pool temperature, however, have a weak effect on the maximum amplitude. Oscillation frequency is decreased with an increase of the air mass in the range of 2.5-4.5 Hz

HUMAN RELIABILITY ANALYSIS FOR COMPUTERIZED PROCEDURES, PART TWO: APPLICABILITY OF CURRENT METHODS

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; David I. Gertman

2012-10-01

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no U.S. nuclear power plant has implemented CPs in its main control room. Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

A toolkit for computerized operating procedure of complex industrial systems with IVI-COM technology

International Nuclear Information System (INIS)

Zhou Yangping; Dong Yujie; Huang Xiaojing; Ye Jingliang; Yoshikawa, Hidekazu

2013-01-01

A human interface toolkit is proposed to help the user develop computerized operating procedure of complex industrial system such as Nuclear Power Plants (NPPs). Coupled with a friendly graphical interface, this integrated tool includes a database, a procedure editor and a procedure executor. A three layer hierarchy is adopted to express the complexity of operating procedure, which includes mission, process and node. There are 10 kinds of node: entrance, exit, hint, manual input, detector, actuator, data treatment, branch, judgment and plug-in. The computerized operating procedure will sense and actuate the actual industrial systems with the interface based on IVI-COM (Interchangeable Virtual Instrumentation-Component Object Model) technology. A prototype system of this human interface toolkit has been applied to develop a simple computerized operating procedure for a simulated NPP. (author)

Procedures for Selecting Items for Computerized Adaptive Tests.

Science.gov (United States)

Kingsbury, G. Gage; Zara, Anthony R.

1989-01-01

Several classical approaches and alternative approaches to item selection for computerized adaptive testing (CAT) are reviewed and compared. The study also describes procedures for constrained CAT that may be added to classical item selection approaches to allow them to be used for applied testing. (TJH)

Development of an improved installation procedure and schedule of RVI modularization for APR1400

International Nuclear Information System (INIS)

Ko, Do Young

2011-01-01

The construction technology for reactor vessel internals (RVI) modularization is one of the most important factors to be considered in reducing the construction period of nuclear power plants. For RVI modularization, gaps between the reactor vessel (RV) core-stabilizing lug and the core support barrel (CSB) snubber lug must be measured using a remote method from outside the RV. In order to measure RVI gaps remotely at nuclear power plant construction sites, certain core technologies must be developed and verified. These include a remote measurement system to measure the gaps between the RV core-stabilizing lug and the CSB snubber lug, an RVI mockup to perform the gap measurement tests, and a new procedure and schedule for RVI installation. A remote measurement system was developed previously, and a gap measurement test was completed successfully using the RVI mockup. We also developed a new procedure and schedule for RVI installation. This paper presents the new and improved installation procedure and schedule for RVI modularization. These are expected to become core technologies that will allow us to shorten the construction period by a minimum of two months compared to the existing installation procedure and schedule

Consideration on hydrogen explosion scenario in APR 1400 containment building during small breakup loss of coolant accident

Energy Technology Data Exchange (ETDEWEB)

Park, Kweonha, E-mail: khpark@kmou.ac.kr [Division of Mechanical & Energy Systems Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of); Khor, Chong Lee, E-mail: itachi_829@hotmail.com [Department of Mechanical Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of)

2015-11-15

Highlights: • Hydrogen behavior in the containment building of APR1400 nuclear plant up to 15 h after the failure happened. • The risk of hydrogen explosion largely depends on the combination of air, hydrogen and steam in the containment. • Hydrogen explosion risk at different locations in the containment was analyzed. - Abstract: This paper describes the analytical result of the potential risk of hydrogen gas up to 15 h after the failure takes place. The major cause of the disaster occurred in Fukushima Daiichi nuclear reactor was the detonation of accumulated hydrogen in the containment by highly increased reactor core temperatures after the failure of the emergency cooling system. The hydrogen risk should be considered in severe accident strategies in current and future NPPs. A hydrogen explosion scenario is proposed. Hydrogen is accumulated on top of the dome during the hydrogen release period. At this point, there are no risk of explosion due to the steam that resides in upper part of the dome. As the hydrogen concentration increase, substantial amount of steams are released. Subsequently, hydrogen is forced into the lower part of the building with high air density—small explosion and dormant steam condensation phase are possible. The light hydrogen rises up slowly with air, gathering on top of the building with high air density. Massive hydrogen explosion is anticipated upon ignition at this stage.

A study on the hydrogen behavior and its mitigation in the APR1400 containment during a severe accident

International Nuclear Information System (INIS)

Kim, Jong Tae; Hong, Seong Wan; Park, Rae Joon; Kim, Sang Baik

2005-02-01

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In this study, the analysis of the hydrogen and steam behavior during selected severe accidents in the APR1400 containment has been conducted by using the GASFLOW code. For the SBLOCA, hydrogen was accumulated in the containment dome region quickly when only PARSs were used. When the igniters were turned on, a standing flame was formed around a coolant pump and burnt most of the hydrogen blown-out. For the TLOFW accident, the flap-type pressure damper installed at the IRWST vents strongly affected the flow structure of the hydrogen. And by the steam-rich and oxygen starvation conditions in the IRWST, DDT is not likely to occur. For the SBO accident, dry hydrogen was release in the IRWST by the assumption of full condensation of the released steam in the IRWST water. In this case, the possibility of flame acceleration is high in the IRWST and annular compartment. In this study two design modifications were proposed in view of the hydrogen mitigation strategy and their effectiveness was evaluated by the GASFLOW analysis

Respirators: APR Issuer Self Study 33461

Energy Technology Data Exchange (ETDEWEB)

Chochoms, Michael [Los Alamos National Laboratory

2016-07-13

Respirators: APR Issuer Self-Study (course 33461) is designed to introduce and familiarize employees selected as air-purifying respirator (APR) issuers at Los Alamos National Laboratory (LANL) with the responsibilities, limitations, procedures, and resources for issuing APRs at LANL. The goal is to enable these issuers to consistently provide proper, functioning APRs to authorized users

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

State Token Petri Net modeling method for formal verification of computerized procedure including operator's interruptions of procedure execution flow

International Nuclear Information System (INIS)

Kim, Yun Goo; Seong, Poong Hyun

2012-01-01

The Computerized Procedure System (CPS) is one of the primary operating support systems in the digital Main Control Room. The CPS displays procedure on the computer screen in the form of a flow chart, and displays plant operating information along with procedure instructions. It also supports operator decision making by providing a system decision. A procedure flow should be correct and reliable, as an error would lead to operator misjudgement and inadequate control. In this paper we present a modeling for the CPS that enables formal verification based on Petri nets. The proposed State Token Petri Nets (STPN) also support modeling of a procedure flow that has various interruptions by the operator, according to the plant condition. STPN modeling is compared with Coloured Petri net when they are applied to Emergency Operating Computerized Procedure. A converting program for Computerized Procedure (CP) to STPN has been also developed. The formal verification and validation methods of CP with STPN increase the safety of a nuclear power plant and provide digital quality assurance means that are needed when the role and function of the CPS is increasing.

Development of a system for a linked analysis of RELAP5 and MAAP4 and its application for an analysis of LB-LOCA case of APR1400

International Nuclear Information System (INIS)

Park, Chang Hwan; Lee, Un Chul; Park, Goon Cherl; Suh, Kune Yull

2004-01-01

The RELAP5 (1995) and MAAP4 (1994) linked-analysis system was designed and illustrative calculation was performed. A large-break loss-of-coolant accident (LBLOCA) was taken as the reference case for the APR1400 (Advanced Power Reactor 1400MWe). For the early phase of this case, the calculational results of two codes have some deviations in water level depletion and hot assembly temperature. Ordinarily, it is considered that RELAP5 has enough accuracy in calculation of the thermal hydraulic behavior of typical PWR during design basis accidents. If the data set for the thermal hydraulic state of RELAP5 can be well-transferred to MAAP4 as an initial condition, the overall transients given by the linked analysis can get more reliability. In this study, the linked analysis system of RELAP5 and MAAP4 doesn't mean the mechanically integrated code structure. The objective of this study is to formulate the linked analysis system of RELAP5 and MAAP4, which should precede construction of mechanically integrated analyzer. Thus, the main scope of this work covers development of the methodology for data linkage and decision of the transfer timing

Tight integration of computerized procedures with plant information at the South Texas Project

International Nuclear Information System (INIS)

Brtis, J.S.; Green, T.

1996-01-01

This paper describes a unique undertaking that is underway at Houston Lighting and Power's South Texas Project (STP). The paper presents an information upgrade project that uses expert system technologies to computerize design change procedures and to tightly integrate the resulting on-line, interactive procedures with the on-line information that design change activities use and generate. This effort will show how procedure computerization can leverage the large investments in plant data. The expected benefits include reduced costs and improved quality of design change work, plus a significant reduction in the burden of configuration management that comes from design changes. Both process computerization and the integration of process with data are being implemented at STP. This work is part of a major migration of information from a mainframe to a LAN platform. This paper will be of greatest interest to those involved in: (1) configuration management, (2) coordinating information to support design change procedures, (3) plant information management, and (4) business process reengineering

Review on the NEI Methodology of Debris Transport Analysis in Sump Blockage Issue for APR1400

International Nuclear Information System (INIS)

Kim, Jong Uk; Lee, Jeong Ik; Hong, Soon Joon; Lee, Byung Chul; Bang, Young Seok

2007-01-01

Since USNRC (United State Nuclear Regulatory Committee) initially addressed post-accident sump performance under Unresolved Safety Issue USI A-43, sump blockage issue has gone through GSI-191, Regulation Guide 1.82, Rev. 3 (RG. 1.82 Rev.3), and generic Letter 2004-02 for PWRs (Pressurized Water Reactors). As a response of these USNRC's activities, NEI 04-07 was issued in order to evaluate the post-accident performance of a plant's recirculation sump. The baseline methodology of NEI 04-07 is composed of break selection, debris generation, latent debris, debris transport, and head loss. In analytical refinement of NEI 04-07, computational fluid dynamic (CFD) is suggested for the evaluation of debris transport in emergency core cooling (ECC) recirculation mode as guided by RG. 1.82 Rev.3. In Korea nuclear industry also keeps step with international activities of this safety issue, with Kori 1 plant as a pioneering edge. Korean nuclear industry has been also pursuing development of an advanced PWR of APR1400, which incorporates several improved safety features. One of the key features, considering sump blockage issue, is the adoption of IRWST (In-containment Refueling Water Storage Tank). This device, as the acronym implies, changes the emergency core cooling water injection pattern. This fact makes us to review the applicability of NEI 04-07's methodology. In this paper we discuss the applicability of NEI 04- 07's methodology, and more over, new methodology is proposed. And finally the preliminary debris transport is analyzed

Core on-line monitoring and computerized procedures systems

International Nuclear Information System (INIS)

Gangloff, W.C.

1986-01-01

The availability of operating nuclear power plants has been affected significantly by the difficulty people have in coping with the complexity of the plants and the operating procedures. Two ways to use modern computer technology to ease the burden of coping are discussed in this paper, an on-line core monitoring system with predictive capability and a computerized procedures system using live plant data. These systems reduce human errors by presenting information rather than simply data, using the computer to manipulate the data, but leaving the decisions to the plant operator

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Evaluation of the Planned Outage Durations in EU-APR

Energy Technology Data Exchange (ETDEWEB)

Jung, Byung Joon; Lee, Keun Sung [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

EU-APR has been designed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. And it is modified and improved from its original design of APR1400. The whole duration varies depending on items for additional process. Refueling and regular maintenance outage is comprised of basic processes and Main turbine-generator outage includes dismantling inspection of main generator and high pressure turbine as a critical path in addition to basic processes. In-Service Inspection Outage includes Automatic ultrasonic inspection on the upper side/lower side of a nuclear reactor as a critical path in addition to basic processes. The planned outage durations of EU-APR are optimized according to the above results. And they are complied with EUR Requirement (EUR 2.2.7.2.2 B), respectively. In addition, outage duration can be reduced with improved operating technology and more maintenance friendly environment including betterment of filling, drain and ventilation.

Evaluation of the Planned Outage Durations in EU-APR

International Nuclear Information System (INIS)

Jung, Byung Joon; Lee, Keun Sung

2016-01-01

EU-APR has been designed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. And it is modified and improved from its original design of APR1400. The whole duration varies depending on items for additional process. Refueling and regular maintenance outage is comprised of basic processes and Main turbine-generator outage includes dismantling inspection of main generator and high pressure turbine as a critical path in addition to basic processes. In-Service Inspection Outage includes Automatic ultrasonic inspection on the upper side/lower side of a nuclear reactor as a critical path in addition to basic processes. The planned outage durations of EU-APR are optimized according to the above results. And they are complied with EUR Requirement (EUR 2.2.7.2.2 B), respectively. In addition, outage duration can be reduced with improved operating technology and more maintenance friendly environment including betterment of filling, drain and ventilation

REVIEW Of COMPUTERIZED PROCEDURE GUIDELINES FOR NUCLEAR POWER PLANT CONTROL ROOMS

Energy Technology Data Exchange (ETDEWEB)

David I Gertman; Katya Le Blanc; Ronald L Boring

2011-09-01

Computerized procedures (CPs) are recognized as an emerging alternative to paper-based procedures for supporting control room operators in nuclear power plants undergoing life extension and in the concept of operations for advanced reactor designs. CPs potentially reduce operator workload, yield increases in efficiency, and provide for greater resilience. Yet, CPs may also adversely impact human and plant performance if not designed and implemented properly. Therefore, it is important to ensure that existing guidance is sufficient to provide for proper implementation and monitoring of CPs. In this paper, human performance issues were identified based on a review of the behavioral science literature, research on computerized procedures in nuclear and other industries, and a review of industry experience with CPs. The review of human performance issues led to the identification of a number of technical gaps in available guidance sources. To address some of the gaps, we developed 13 supplemental guidelines to support design and safety. This paper presents these guidelines and the case for further research.

Application of computerized procedure system for AP1000

International Nuclear Information System (INIS)

Wen Fang

2012-01-01

With the tendency of digitalisation in instrumentation and control system of nuclear power plants, AP1000, as well as other advanced PWRs, is provided with the technical basis for the application of CPS (computerized procedure system). This paper makes a brief introduction on CPS construction and function. CPS, as an advanced procedure system, does not only have the function of electronic indication for operation procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. In addition, advantages of CPS compared with paper-based procedures and comparison with digital operating system of one M310 plus advanced nuclear power plant are described. Moreover, based on current situation, this paper offers several suggestions on CPS localization for Sanmen AP1000 nuclear power project. Besides, the last part of this paper discusses problems we might meet during the process of CPS localization. (author)

Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

International Nuclear Information System (INIS)

Bang, Young Seok

2015-01-01

The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage

Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

Energy Technology Data Exchange (ETDEWEB)

Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage.

Extension of Surveillance Test Interval of Safety Injection Pump for APR-1400 Reactors to Improve Reliability and Availability of the Pump

Energy Technology Data Exchange (ETDEWEB)

Osama, A. Rezk; Jung, J. C.; Lee, Yong-Kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

The safety features function to localize, control, mitigate, and terminate such incidents and to hold exposure levels below applicable limits. The safety injection system is comprised of four independent mechanical trains without any tie line among the injection paths and two electrical divisions. Each train has one active Safety Injection Pump (SIP) and one passive Safety Injection Tank (SIT) equipped with a Fluidic Device (FD), each train provides 50% of the minimum injection flow rate for breaks larger than the size of a direct vessel injection line. For breaks equal to or smaller than the size of a direct vessel injection line, each train has 100% of the required capacity. The low pressure injection pumps with common header installed in the conventional design are eliminated, and the functions for safety injection and shutdown cooling are separated. The arrangement of safety injection system for APR-1400 as shown in figure (1). The results obtained in this work show that STI extensions for the SIS feasible without any unacceptable increase in the plant total risk, STI extensions are acceptable for safety injection system to provide plant operational flexibility in the performance of both corrective and preventive maintenance for the safety injection system.

A Computerized Procedure linked to Virtual Equipment

International Nuclear Information System (INIS)

Jung, Yeon Sub; Song, Tae Young

2011-01-01

Digital, information, and communication technologies have change human's behavior. This is because human has limitation to memorize and process information. Human has to access other information and real time information for important decisions. Those technologies are playing important roles. Nuclear power plants cannot be exception. Many accidents in nuclear power plants result from absent or incorrect information. The information for nuclear personnel is context sensitive. They don't have enough time to verify the context sensitive information. Therefore they skip the information, as resulting in incident. Nuclear personnel are usually carrying static paper procedures during local task performance. The procedure guides them steps to follow. There is, however, no dynamic and context sensitive information in the paper. The effect of the work is evaluated once while getting permission of the work. Afterward they are not informed. The static paper is generally simplified, so that it does not show detail of equipment being manipulated. Particularly novice workers feel difficult to understand the procedure due to lack of detail. Pictures of equipment inserted in the procedure are not enough for comprehension. A computerized procedure linked with virtual equipment is one of the best solutions to increase the detail of procedure. Virtual equipment, however, has still limitation not to provide real time information, because the virtual equipment is not synchronized with real plants

An ergonomics study of computerized emergency operating procedures: Presentation style, task complexity, and training level

International Nuclear Information System (INIS)

Xu Song; Song Fei; Li Zhizhong; Zhao Qianyi; Luo Wei; He Xuhong; Salvendy, Gavriel

2008-01-01

Emergency operating procedures (EOPs) are widely used in nuclear power plants (NPPs). With the development of information technology, computerized EOPs are taking the place of paper-based ones. Unlike paper-based EOPs, the industrial practice of computerized EOPs is still quite limited. Ergonomics issues of computerized EOPs have not been studied adequately. This study focuses on the effects of EOP presentation style, task complexity, and training level on the performance of the operators in the execution of computerized EOPs. One simulated computerized EOP system was developed to present two EOPs, each with different task complexity levels, by two presentation styles (Style A: one- and two-dimensional flowcharts combination; Style B: two-dimensional flowchart and success logic tree combination). Forty subjects participated in the experiment of EOP execution using the simulated system. Statistical analysis of the experimental results indicates that: (1) complexity, presentation style, and training level all can significantly influence the error rate. High-complexity tasks and lack of sufficient training may lead to a higher error rate. Style B caused a significantly higher error rate than style A did in the skilled phase. So the designers of computerized procedures should take the presentation styles of EOPs into account. (2) Task complexity and training level can significantly influence operation time. No significant difference was found in operation time between the two presentation styles. (3) Training level can also significantly influence the subjective workload of EOPs operations. This implies that adequate training is very important for the performance of computerized EOPs even if emergency responses with computerized EOPs are much more simple and easy than that with paper-based EOPs

An ergonomics study of computerized emergency operating procedures: Presentation style, task complexity, and training level

Energy Technology Data Exchange (ETDEWEB)

Xu Song; Song Fei [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China); Li Zhizhong [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China)], E-mail: zzli@tsinghua.edu.cn; Zhao Qianyi; Luo Wei [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China); He Xuhong [Scanpower Risk Management China Inc., Towercrest International Plaza, No. 3 Maizidian West Road, Chaoyang District, Beijing 100016 (China); Salvendy, Gavriel [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China)

2008-10-15

Emergency operating procedures (EOPs) are widely used in nuclear power plants (NPPs). With the development of information technology, computerized EOPs are taking the place of paper-based ones. Unlike paper-based EOPs, the industrial practice of computerized EOPs is still quite limited. Ergonomics issues of computerized EOPs have not been studied adequately. This study focuses on the effects of EOP presentation style, task complexity, and training level on the performance of the operators in the execution of computerized EOPs. One simulated computerized EOP system was developed to present two EOPs, each with different task complexity levels, by two presentation styles (Style A: one- and two-dimensional flowcharts combination; Style B: two-dimensional flowchart and success logic tree combination). Forty subjects participated in the experiment of EOP execution using the simulated system. Statistical analysis of the experimental results indicates that: (1) complexity, presentation style, and training level all can significantly influence the error rate. High-complexity tasks and lack of sufficient training may lead to a higher error rate. Style B caused a significantly higher error rate than style A did in the skilled phase. So the designers of computerized procedures should take the presentation styles of EOPs into account. (2) Task complexity and training level can significantly influence operation time. No significant difference was found in operation time between the two presentation styles. (3) Training level can also significantly influence the subjective workload of EOPs operations. This implies that adequate training is very important for the performance of computerized EOPs even if emergency responses with computerized EOPs are much more simple and easy than that with paper-based EOPs.

Radiological Impacts Assessment during Normal Decommissioning Operation for EU-APR

Energy Technology Data Exchange (ETDEWEB)

Hwang, Do Hyun; Lee, Keun Sung [KHNP CRI, Daejeon (Korea, Republic of); Lee, ChongHui [KEPCO Engineering and Construction, Gimcheon (Korea, Republic of)

2016-10-15

In this paper, radiological impacts on human beings during normal execution of the decommissioning operations from the current standard design of EU-APR which has been modified and improved from its original design of APR1400 to comply with EUR, are evaluated. Decommissioning is the final phase in the life cycle of a nuclear installation, covering all activities from shutdown and removal of fissile material to environmental restoration of the site. According to article 5.4 specified in chapter 2.20 of European Utility Requirements (EUR), all relevant radiological impacts on human being should be considered during the environmental assessment of decommissioning, including external exposure from direct radiation of plant and other radiation sources, and internal exposure due to inhalation and ingestion. In this paper, radiological impacts on human beings during normal circumstances of the decommissioning operation were evaluated from the current standard design of EU-APR based on the simple transport model and practical generic methodology for assessing the radiological impact provided by IAEA. The results of dose assessment fulfilled the dose limit for all scenarios.

Analysis of a molten pool natural convection in the APR1400 RPV at a severe accident

International Nuclear Information System (INIS)

Kim, Jong Tae; Park, Rae Joon; Kim, Sang Baik

2005-01-01

During a hypothetical severe accident, reactor fuel rods and structures supporting them are melted and relocated in the lower head of the reactor vessel. These relocated molten materials could be separated by their density difference and construct metal/oxide stratified pools in the lower head. A decay heat generated from the fuel material is transferred to the vessel wall and upper structures remaining in the reactor vessel by natural convection. As shown in Fig. 1 two-layered stratified molten pool is developed in the reactor lower vessel. The oxidic pool usually constructed by the mixture of uranium oxide and zirconium oxide. The melting temperature of the oxidic material is very high compared to the steel vessel and metallic layer. And highly turbulent natural convection generated by the decay heat enhances heat transfer to the boundary of the oxidic pool. By this thermal mechanism, oxide curst is developed around the oxidic layer as shown in Fig. 1. The oxidic pool is bounded thermally and fluid-dynamically by the developed crust. By this boundedness, the heat transfer structure in the stratified oxidic/metallic pool can be solved separately. The thermal boundary condition of the oxidic pool is isothermal with constant melting temperature of the oxidic material. The decay heat is transfer to side wall and upper interface between oxidic and metallic layer. Turbulent natural convection is dominant heat transfer mechanism in the oxidic pool. The heat transferred from the bottom oxidic layer is imposed to the upper metallic layer. This transferred heat in the metallic pool is removed through side and upper surface, which is augmented also by natural convection developed in the pool. In this study, a molten pool natural convection in the APR1400 RPV during a severe accident is simulated using the Lilac code and the calculated heat flux distribution on the reactor vessel wall is compared with a lumped-parameter (LP) prediction

A Procedure for the Computerized Analysis of Cleft Palate Speech Transcription

Science.gov (United States)

Fitzsimons, David A.; Jones, David L.; Barton, Belinda; North, Kathryn N.

2012-01-01

The phonetic symbols used by speech-language pathologists to transcribe speech contain underlying hexadecimal values used by computers to correctly display and process transcription data. This study aimed to develop a procedure to utilise these values as the basis for subsequent computerized analysis of cleft palate speech. A computer keyboard…

Development of an amphibious robot for visual inspection of APR1400 Npp IRWST strainer

International Nuclear Information System (INIS)

Jang, You Hyun; Kim, Jong Seog

2014-01-01

An amphibious inspection robot system (hereafter AIROS) is being developed to visually inspect the in-containment refueling storage water tank (hereafter IRWST) strainer in APR1400 instead of a human diver. Four IRWST strainers are located in the IRWST, which is filled with boric acid water. Each strainer has 108 sub-assembly strainer fin modules that should be inspected with the VT-3 method according to Reg. guide 1.82 and the operation manual. AIROS has 6 thrusters for submarine voyage and 4 legs for walking on the top of the strainer. An inverse kinematic algorithm was implemented in the robot controller for exact walking on the top of the IRWST strainer. The IRWST strainer has several top cross braces that are extruded on the top of the strainer, which can be obstacles of walking on the strainer, to maintain the frame of the strainer. Therefore, a robot leg should arrive at the position beside the top cross brace. For this reason, we used an image processing technique to find the top cross brace in the sole camera image. The sole camera image is processed to find the existence of the top cross brace using the cross edge detection algorithm in real time. A 5-DOF robot arm that has multiple camera modules for simultaneous inspection of both sides can penetrate narrow gaps. For intuitive presentation of inspection results and for management of inspection data, inspection images are stored in the control PC with camera angles and positions to synthesize and merge the images. The synthesized images are then mapped in a 3D CAD model of the IRWST strainer with the location information. An IRWST strainer mock-up was fabricated to teach the robot arm scanning and gaiting. It is important to arrive at the designated position for inserting the robot arm into all of the gaps. Exact position control without anchor under the water is not easy. Therefore, we designed the multi leg robot for the role of anchoring and positioning. Quadruped robot design of installing sole

Development of an amphibious robot for visual inspection of APR1400 Npp IRWST strainer

Energy Technology Data Exchange (ETDEWEB)

Jang, You Hyun; Kim, Jong Seog [Korea Hydro Nuclear Power Central Research Institute, Daejeon (Korea, Republic of)

2014-06-15

An amphibious inspection robot system (hereafter AIROS) is being developed to visually inspect the in-containment refueling storage water tank (hereafter IRWST) strainer in APR1400 instead of a human diver. Four IRWST strainers are located in the IRWST, which is filled with boric acid water. Each strainer has 108 sub-assembly strainer fin modules that should be inspected with the VT-3 method according to Reg. guide 1.82 and the operation manual. AIROS has 6 thrusters for submarine voyage and 4 legs for walking on the top of the strainer. An inverse kinematic algorithm was implemented in the robot controller for exact walking on the top of the IRWST strainer. The IRWST strainer has several top cross braces that are extruded on the top of the strainer, which can be obstacles of walking on the strainer, to maintain the frame of the strainer. Therefore, a robot leg should arrive at the position beside the top cross brace. For this reason, we used an image processing technique to find the top cross brace in the sole camera image. The sole camera image is processed to find the existence of the top cross brace using the cross edge detection algorithm in real time. A 5-DOF robot arm that has multiple camera modules for simultaneous inspection of both sides can penetrate narrow gaps. For intuitive presentation of inspection results and for management of inspection data, inspection images are stored in the control PC with camera angles and positions to synthesize and merge the images. The synthesized images are then mapped in a 3D CAD model of the IRWST strainer with the location information. An IRWST strainer mock-up was fabricated to teach the robot arm scanning and gaiting. It is important to arrive at the designated position for inserting the robot arm into all of the gaps. Exact position control without anchor under the water is not easy. Therefore, we designed the multi leg robot for the role of anchoring and positioning. Quadruped robot design of installing sole

10 CFR 2.1400 - Purpose and scope of subpart N.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Purpose and scope of subpart N. 2.1400 Section 2.1400... ORDERS Expedited Proceedings with Oral Hearings § 2.1400 Purpose and scope of subpart N. The purpose of... parties otherwise agree and request the application of subpart N procedures, and proceedings for the...

DEVELOPMENT OF AN AMPHIBIOUS ROBOT FOR VISUAL INSPECTION OF APR1400 NPP IRWST STRAINER ASSEMBLY

Directory of Open Access Journals (Sweden)

YOU HYUN JANG

2014-06-01

Full Text Available An amphibious inspection robot system (hereafter AIROS is being developed to visually inspect the in-containment refueling storage water tank (hereafter IRWST strainer in APR1400 instead of a human diver. Four IRWST strainers are located in the IRWST, which is filled with boric acid water. Each strainer has 108 sub-assembly strainer fin modules that should be inspected with the VT-3 method according to Reg. guide 1.82 and the operation manual. AIROS has 6 thrusters for submarine voyage and 4 legs for walking on the top of the strainer. An inverse kinematic algorithm was implemented in the robot controller for exact walking on the top of the IRWST strainer. The IRWST strainer has several top cross braces that are extruded on the top of the strainer, which can be obstacles of walking on the strainer, to maintain the frame of the strainer. Therefore, a robot leg should arrive at the position beside the top cross brace. For this reason, we used an image processing technique to find the top cross brace in the sole camera image. The sole camera image is processed to find the existence of the top cross brace using the cross edge detection algorithm in real time. A 5-DOF robot arm that has multiple camera modules for simultaneous inspection of both sides can penetrate narrow gaps. For intuitive presentation of inspection results and for management of inspection data, inspection images are stored in the control PC with camera angles and positions to synthesize and merge the images. The synthesized images are then mapped in a 3D CAD model of the IRWST strainer with the location information. An IRWST strainer mock-up was fabricated to teach the robot arm scanning and gaiting. It is important to arrive at the designated position for inserting the robot arm into all of the gaps. Exact position control without anchor under the water is not easy. Therefore, we designed the multi leg robot for the role of anchoring and positioning. Quadruped robot design of

A production system for computerized emergency procedures tracking

International Nuclear Information System (INIS)

Petrick, W.; Ng, K.; Stuart, C.; Cain, D.

1988-01-01

This paper describes a methodology used to implement a computerized Emergency Operating Procedures (EOP) tracking system based on artificial intelligence (AI) technologies. In AI parlance, the system is a production system in which the knowledge representation is a set of productions (IF-THEN rules) which are interpreted by an inference engine or rule interpreter. The inference engine looks for pattern matches in the rule premises which lead to a conclusion. This inferred conclusion is subsequently used in further pattern matches. By separating the knowledge representation from the inference engine, the system can readily accommodate changes to the written procedures. The system is written in the C language and is co-resident on the same computer as the Safety Parameter Display System (SPDS). The plant EOPs have been written in about 250 rules which are evaluated in less than one second using about 10-20% of the CPU. The conclusions are available to multiple users within seconds after a parameter value changes. The benefits achieved with this production rule system include compatibility with the written procedures used as the knowledge source, compatibility with control room instrumentation, compatibility with the SPDS, realtime performance, and efficient computer usage

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

Analysis of risk in computerized tomography and other diagnostic radiology procedures

International Nuclear Information System (INIS)

Mossman, K.L.

1982-01-01

Medical practice entails continuous risks to the patient taken in good faith by the physician for the benefit of the patient. Risk of radiation induced cancer death approximates 10(-4) per cGy (rad). Assuming an average whole body dose of 0.1 cGy for many diagnostic X-ray procedures, the probability of radiation-induced cancer death is about 10(-5). The purpose of this paper is to compare the risks of common diagnostic X-ray procedures including computerized tomography (CT) with risks of smoking or automobile travel. Such comparisons should be constructive in putting radiation in perspective and facilitating explanation of risk/benefit to patients

A Comparison of Procedures for Content-Sensitive Item Selection in Computerized Adaptive Tests.

Science.gov (United States)

Kingsbury, G. Gage; Zara, Anthony R.

1991-01-01

This simulation investigated two procedures that reduce differences between paper-and-pencil testing and computerized adaptive testing (CAT) by making CAT content sensitive. Results indicate that the price in terms of additional test items of using constrained CAT for content balancing is much smaller than that of using testlets. (SLD)

Feasibility study of computerized presentation of procedures of a nuclear plant

International Nuclear Information System (INIS)

Takagawa, Kenichi; Niwa, Yuji; Urakami, Masahiro

2010-01-01

Introduction of a digital instrumentation control system is in progress in recent nuclear power plants, and an application of Computer Based Procedures (CBP) is being considered. The Institute of Nuclear Safety System, Incorporated has developed a system for a Computerized Presentation of Procedures (CPP) which corresponds to CBP. By fiscal year 2002, all items in the emergency operation procedures of a model plant were made displayable on the CPP. In 2003, the CPP instead of the paper based procedures was used for trial by the operators of a nuclear power plant in an actual training simulator, and feasibility of its practical application was evaluated. Based on the result, we considered subjects which should be taken into account in the future. As a result, improvements of CPP and its navigation system, etc. to enable operations with multi-procedures in parallel were extracted as primary subjects for introduction of CBP in the future as well as practical application of CPP. (author)

Integrated System Validation Usability Questionnaire: Computerized Procedures

International Nuclear Information System (INIS)

Garcés, Ma. I.; Torralba, B.

2015-01-01

The Research and Development (R&D) project on “Theoretical and Methodological Approaches to Integrated System Validation of Control Rooms, 2014-2015”, in which the research activities described in this report are framed, has two main objectives: to develop the items for an usability methodology conceived as a part of the measurement framework for performance-based control room evaluation that the OECD Halden Reactor Project will test in the experiments planned for 2015; and the statistical analysis of the data generated in the experimental activities of the Halden Man-Machine Laboratory (HAMMLAB) facility, with previous usability questionnaires, in 2010 and 2011. In this report, the procedure designed to meet the first goal of the project is described, in particular, the process followed to identify the items related to operating procedures, both computer and paper-based, one of the elements to be included in the usability questionnaire. Three phases are performed, in the first one, the approaches developed by the United States Nuclear Regulatory Commission, NRC, are reviewed, the models used by the nuclear industry and their technical support organizations, mainly, the Electric Power Research Institute, EPRI, are analyzed, and scientist advances are also explored. In the remaining stages, general and specific guidelines for computerized and paper-based procedures are compared and criteria for the preliminary selection of the items that should be incorporated into the usability questionnaire are defined. This proposal will be reviewed and adapted by the Halden Reactor Project to the design of the specific experiments performed in HAMLAB.

Assessing mental workload and situation awareness in the evaluation of computerized procedures in the main control room

Energy Technology Data Exchange (ETDEWEB)

Yang, Chih-Wei, E-mail: yangcw@iner.gov.tw [Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yang, Li-Chen; Cheng, Tsung-Chieh [Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Jou, Yung-Tsan; Chiou, Shian-Wei [Department of Industrial Engineering, Chung-Yuan Christian University, 200, Chung Pei Rd., Chung-Li 32023, Taiwan (China)

2012-09-15

Highlights: Black-Right-Pointing-Pointer This study investigates procedure types' effects on operators' performance. Black-Right-Pointing-Pointer The computer-based procedure is suggested to be implemented in the main control room. Black-Right-Pointing-Pointer The computer-based procedure brings lowest mental workload. Black-Right-Pointing-Pointer And it also generates fewer error of omission, and the highest situation awareness. Black-Right-Pointing-Pointer The shift supervisor has the highest workload and the lowest situation awareness. - Abstract: Computerized procedure (CP) system has been developed in nuclear power plant (NPP) instrumentation and control (I and C) system. The system may include normal operating procedures (OPs), abnormal operating procedures (AOPs), alarm response procedures (ARPs), surveillance test procedures (STPs) and/or emergency operating procedures (EOPs). While there are many ways to evaluate computerized procedures design, the user's mental workload and situation awareness (SA) are particularly important considerations in the supervisory control of safety-critical systems. Users' mental workload and situation awareness may be influenced by human factor issues relating to computerized procedures, e.g., level of automation, dealing with (partially) unavailable I and C, switching to back-up system (e.g., paper-based procedures). Some of the positive impacts of CPs on operator performance include the following: tasks can be performed more quickly; overall workload can be reduced; cognitive workload can be minimized; fewer errors may be made in transitioning through or between procedures. However, various challenges have also been identified with CP systems. These should be addressed in the design and implementation of CPs where they are applicable. For example, narrower 'field of view' provided by CP systems than with paper-based procedures could reduce crew communications and crewmember awareness of the

Assessing mental workload and situation awareness in the evaluation of computerized procedures in the main control room

International Nuclear Information System (INIS)

Yang, Chih-Wei; Yang, Li-Chen; Cheng, Tsung-Chieh; Jou, Yung-Tsan; Chiou, Shian-Wei

2012-01-01

Highlights: ► This study investigates procedure types’ effects on operators’ performance. ► The computer-based procedure is suggested to be implemented in the main control room. ► The computer-based procedure brings lowest mental workload. ► And it also generates fewer error of omission, and the highest situation awareness. ► The shift supervisor has the highest workload and the lowest situation awareness. - Abstract: Computerized procedure (CP) system has been developed in nuclear power plant (NPP) instrumentation and control (I and C) system. The system may include normal operating procedures (OPs), abnormal operating procedures (AOPs), alarm response procedures (ARPs), surveillance test procedures (STPs) and/or emergency operating procedures (EOPs). While there are many ways to evaluate computerized procedures design, the user's mental workload and situation awareness (SA) are particularly important considerations in the supervisory control of safety-critical systems. Users’ mental workload and situation awareness may be influenced by human factor issues relating to computerized procedures, e.g., level of automation, dealing with (partially) unavailable I and C, switching to back-up system (e.g., paper-based procedures). Some of the positive impacts of CPs on operator performance include the following: tasks can be performed more quickly; overall workload can be reduced; cognitive workload can be minimized; fewer errors may be made in transitioning through or between procedures. However, various challenges have also been identified with CP systems. These should be addressed in the design and implementation of CPs where they are applicable. For example, narrower “field of view” provided by CP systems than with paper-based procedures could reduce crew communications and crewmember awareness of the status and progress through the procedure. Based on a human factors experiment in which each participant monitored and controlled multiple simulated

Aiming for a fully integrated computerized procedure system

International Nuclear Information System (INIS)

Marron, J. E.

2006-01-01

A fully integrated Computerized Procedure System must provide, at a minimum, a) Specification: access to design basis procedures, b) Monitoring: incorporation of real-time plant status, c) Advise: highlighting likely decision paths, and d) Reporting: logging conditions and actions taken. The CPS plays a critical role in overcoming the human factors that lead to accidents. At the same time it can be an essential tool in providing the information and automation to augment what humans do best, identify patterns and make associative leaps in the presence of ambiguous data. Timeliner and TaskGuide are examples of CPS that have evolved from projects in the aerospace industry. They illustrate certain common characteristics of a CPS, namely the knowledge base, user interface, and traceability features. The complexity and number of procedures for a current nuclear project has led to the development of two tools, the Power Generation Control System (PGCS) and the Online Procedure System (OLPS). Together, these systems address the knowledge-base and user interface aspects of a CPS and go a long way in addressing other areas. PGCS and OLPS contain full configuration management capabilities for procedures and the operating recipe. They include administrative functions for online and offline management of documents and data. Some lessons learned from this pair of programs developed by Invensys is the need for more integrated recording mechanisms. The future of CPS is likely to see higher integration of the document access, system status, decision support and logging capabilities. The CPS may evolve into the standard operational interface. Internet technologies that are common-place today have made the possibility of the Active Document a reality. The OPC Foundation is pursuing standards that may accelerate such developments. (authors)

A Study on Large Display Panel Design for the Countermeasures against Team Errors within the Main Control Room of APR-1400

Energy Technology Data Exchange (ETDEWEB)

Kim, Sa Kil; Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The personal aspect of human errors has been mainly overcome by virtue of the education and training. However, in the system aspect, the education and training system needs to be reconsidered for more effective reduction of human errors affected from various systems hazards. Traditionally the education and training systems are mainly not focused on team skills such as communication, situational awareness, and coordination, etc. but individual knowledge, skill, and attitude. However, the team factor is one of the crucial issues to reduce the human errors in most industries. In this study, we identify the emerging types of team errors, especially, in digitalized control room of nuclear power plants such as the APR-1400 main control room. Most works in nuclear industry are to be performed by a team of more than two persons. Even though the individual errors can be detected and recovered by the qualified others and/or the well trained team, it is rather seldom that the errors by team could be easily detected and properly recovered by the team itself. Note that the team is defined as two or more people who are appropriately interacting with each other, and the team is a dependent aggregate, which accomplishes a valuable goal. Team error is one of the typical organizational errors that may occur during performing operations in nuclear power plants. The large display panel is a representative feature of digitalized control room. As a group-view display, the large display panel provides plant overview to the operators. However, in terms of team performance and team errors, the large display panel is on a discussion board still because the large display panel was designed just a concept of passive display. In this study, we will propose revised large display panel which is integrated with several alternative interfaces against feasible team errors.

A Study on Large Display Panel Design for the Countermeasures against Team Errors within the Main Control Room of APR-1400

International Nuclear Information System (INIS)

Kim, Sa Kil; Lee, Yong Hee

2015-01-01

The personal aspect of human errors has been mainly overcome by virtue of the education and training. However, in the system aspect, the education and training system needs to be reconsidered for more effective reduction of human errors affected from various systems hazards. Traditionally the education and training systems are mainly not focused on team skills such as communication, situational awareness, and coordination, etc. but individual knowledge, skill, and attitude. However, the team factor is one of the crucial issues to reduce the human errors in most industries. In this study, we identify the emerging types of team errors, especially, in digitalized control room of nuclear power plants such as the APR-1400 main control room. Most works in nuclear industry are to be performed by a team of more than two persons. Even though the individual errors can be detected and recovered by the qualified others and/or the well trained team, it is rather seldom that the errors by team could be easily detected and properly recovered by the team itself. Note that the team is defined as two or more people who are appropriately interacting with each other, and the team is a dependent aggregate, which accomplishes a valuable goal. Team error is one of the typical organizational errors that may occur during performing operations in nuclear power plants. The large display panel is a representative feature of digitalized control room. As a group-view display, the large display panel provides plant overview to the operators. However, in terms of team performance and team errors, the large display panel is on a discussion board still because the large display panel was designed just a concept of passive display. In this study, we will propose revised large display panel which is integrated with several alternative interfaces against feasible team errors

Development of a testing methodology for computerized procedure system based on JUnit framework and MFM

International Nuclear Information System (INIS)

Qin, Wei

2004-02-01

Paper Based Procedure (PBP) and Computerized Procedure System (CPS) are studied to demonstrate that it is necessary to develop CPS in Nuclear Power Plant (NPP) Instrumentation and Control (I and C) system. Computerized procedure system is actually a software system. All the desired and undesired properties of a software system can be described and evaluated as software qualities. Generally, software qualities can be categorized into product quality and process quality. In order to achieve product quality, the process quality of a software system should also be considered and achieved. Characteristics of CPS will be described to analyse the product and process of an example CPS: ImPRO. At the same time, several main product and process issues will be analysed from Verification and Validation (V and V) point of view. It is concluded and suggested that V and V activities can also be regarded as a software development process, this point of view then is applied to the V and V activities of ImPRO as a systematic approach of testing of ImPRO. To support and realize this approach, suitable testing technologies and testing strategies are suggested based on JUnit framework and Multi-level Flow Modeling (MFM)

Influence of step complexity and presentation style on step performance of computerized emergency operating procedures

Energy Technology Data Exchange (ETDEWEB)

Xu Song [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China); Li Zhizhong [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China)], E-mail: zzli@tsinghua.edu.cn; Song Fei; Luo Wei; Zhao Qianyi; Salvendy, Gavriel [Department of Industrial Engineering, Tsinghua University, Beijing 100084 (China)

2009-02-15

With the development of information technology, computerized emergency operating procedures (EOPs) are taking the place of paper-based ones. However, ergonomics issues of computerized EOPs have not been studied adequately since the industrial practice is quite limited yet. This study examined the influence of step complexity and presentation style of EOPs on step performance. A simulated computerized EOP system was developed in two presentation styles: Style A: one- and two-dimensional flowcharts combination; Style B: two-dimensional flowchart and success logic tree combination. Step complexity was quantified by a complexity measure model based on an entropy concept. Forty subjects participated in the experiment of EOP execution using the simulated system. The results of data analysis on the experiment data indicate that step complexity and presentation style could significantly influence step performance (both step error rate and operation time). Regression models were also developed. The regression analysis results imply that operation time of a step could be well predicted by step complexity while step error rate could only partly predicted by it. The result of a questionnaire investigation implies that step error rate was influenced not only by the operation task itself but also by other human factors. These findings may be useful for the design and assessment of computerized EOPs.

Some procedures for computerized ability testing

NARCIS (Netherlands)

van der Linden, Willem J.; Zwarts, Michel A.

1989-01-01

For computerized test systems to be operational, the use of item response theory is a prerequisite. As opposed to classical test theory, in item response models the abilities of the examinees and the properties of the items are parameterized separately. Hence, when measuring the abilities of

A study on a systematic approach of verification and validation of a computerized procedure system: ImPRO

International Nuclear Information System (INIS)

Qin, Wei; Seong, Poong Hyun

2003-01-01

Paper Based Procedure (PBP) and Computerized Procedure System (CPS) are studied to demonstrate that it is necessary to develop CPS in NPP I and C system. Computerized procedure system is actually a software system. All the desired and undesired properties of a software system can be described and evaluated as software qualities. Generally, software qualities can be categorized into product quality and process quality. In order to achieve product quality, the process quality of a software system should also be considered and achieved. Characteristics of CPS will be described to analyse the product and process of an example CPS: ImPRO. At the same time, several main product and process issues will be analysed from Verification and Validation (V and V) point of view. It is concluded and suggested that V and V activities can also be regarded as software development process, this point of view then is applied to the V and V activities of ImPRO as a systematic approach of V and V of ImPRO. To support and realize this approach, suitable testing technologies and testing strategies are suggested

7 CFR 1400.2 - Administration.

Science.gov (United States)

2010-01-01

... 7 Agriculture 10 2010-01-01 2010-01-01 false Administration. 1400.2 Section 1400.2 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT CORPORATION, DEPARTMENT OF... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.2 Administration. (a) The regulations in...

Operational Strategy of CBPs for load balancing of Operators in Advanced Main Control Room

International Nuclear Information System (INIS)

Kim, Seunghwan; Kim, Yochan; Jung, Wondea

2014-01-01

With the using of a computer-based control room in an APR1400 (Advanced Pressurized Reactor-1400), the operators' behaviors in the main control room had changed. However, though the working environment of operators has been changed a great deal, digitalized interfaces can also change the cognitive tasks or activities of operators. First, a shift supervisor (SS) can confirm/check the conduction of the procedures and the execution of actions of board operators (BOs) while confirming directly the operation variables without relying on the BOs. Second, all operators added to their work the use of a new CBP and Soft Controls, increasing their procedural workload. New operational control strategies of CBPs are necessary for load balancing of operator's task load in APR1400. In this paper, we compared the workloads of operators in an APR1400 who work with two different usages of the CBP. They are SS oriented usage and SS-BO collaborative usage. In this research, we evaluated the workloads of operators in an advanced main control room by the COCOA method. Two types of CBP usages were defined and the effects of these usages on the workloads were investigated. The obtained results showed that the workloads between operators in a control room can be balanced according to the CBP usages by assigning control authority to the operators

Operational Strategy of CBPs for load balancing of Operators in Advanced Main Control Room

Energy Technology Data Exchange (ETDEWEB)

Kim, Seunghwan; Kim, Yochan; Jung, Wondea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

With the using of a computer-based control room in an APR1400 (Advanced Pressurized Reactor-1400), the operators' behaviors in the main control room had changed. However, though the working environment of operators has been changed a great deal, digitalized interfaces can also change the cognitive tasks or activities of operators. First, a shift supervisor (SS) can confirm/check the conduction of the procedures and the execution of actions of board operators (BOs) while confirming directly the operation variables without relying on the BOs. Second, all operators added to their work the use of a new CBP and Soft Controls, increasing their procedural workload. New operational control strategies of CBPs are necessary for load balancing of operator's task load in APR1400. In this paper, we compared the workloads of operators in an APR1400 who work with two different usages of the CBP. They are SS oriented usage and SS-BO collaborative usage. In this research, we evaluated the workloads of operators in an advanced main control room by the COCOA method. Two types of CBP usages were defined and the effects of these usages on the workloads were investigated. The obtained results showed that the workloads between operators in a control room can be balanced according to the CBP usages by assigning control authority to the operators.

Challenges to a Man-Machine Interface System in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Jung, Yeon Sub [KHNP, Daejeon (Korea, Republic of)

2009-10-15

When the APR1400 project started nearly 20 years ago, advanced features such as intelligent alarms, computer-based procedures, and soft control technology were introduced. Although some of these technologies had been utilized in fossil plants, APR1400 opened a new horizon in MMIS for nuclear power plants. Currently, however, it appears that further challenges do not exist in the MMIS area. Engineers are simply busy in designing and constructing the next nuclear plant without exploiting the new features of MMIS. This paper explains newly emerging and feasible technology and suggests new ideas for MMIS compared with other industries.

Challenges to a Man-Machine Interface System in Nuclear Power Plants

International Nuclear Information System (INIS)

Jung, Yeon Sub

2009-01-01

When the APR1400 project started nearly 20 years ago, advanced features such as intelligent alarms, computer-based procedures, and soft control technology were introduced. Although some of these technologies had been utilized in fossil plants, APR1400 opened a new horizon in MMIS for nuclear power plants. Currently, however, it appears that further challenges do not exist in the MMIS area. Engineers are simply busy in designing and constructing the next nuclear plant without exploiting the new features of MMIS. This paper explains newly emerging and feasible technology and suggests new ideas for MMIS compared with other industries

Computerized Italian criticality guide, description and validation

International Nuclear Information System (INIS)

Carotenuto, M.; Landeyro, P.A.

1988-10-01

Our group is developing an 'expert system' for collecting engineering know-how on back-end nuclear plant design. An expert system is the most suitable software tool for our problem. During the analysis, the design process was divided into different branches. At each branch of the design process the Expert System relates a computerized design procedure. Any design procedure is composed of a set of design methods, together with their condition of application and reliability limits. In the framework of this expert system, the nuclear criticality safety analysis procedure was developed, in the form of a computerized criticality guide, attempting to reproduce the designer's normal 'reasoning' process. The criticality guide is composed of two parts: A computerized text, including theory, a description of the accidents occurred in the past and a description of the italian design experience; An interactive computer aided calculation module, containing a graphical facility for critical parameter curves. In the present report are presented the criticality guide (computerized Italian Criticality Guide) and its validation test. (author)

Computerized Italian criticality guide, description and validation

Energy Technology Data Exchange (ETDEWEB)

Carotenuto, M; Landeyro, P A [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Casaccia (Italy)

1988-10-15

Our group is developing an 'expert system' for collecting engineering know-how on back-end nuclear plant design. An expert system is the most suitable software tool for our problem. During the analysis, the design process was divided into different branches. At each branch of the design process the Expert System relates a computerized design procedure. Any design procedure is composed of a set of design methods, together with their condition of application and reliability limits. In the framework of this expert system, the nuclear criticality safety analysis procedure was developed, in the form of a computerized criticality guide, attempting to reproduce the designer's normal 'reasoning' process. The criticality guide is composed of two parts: A computerized text, including theory, a description of the accidents occurred in the past and a description of the italian design experience; An interactive computer aided calculation module, containing a graphical facility for critical parameter curves. In the present report are presented the criticality guide (computerized Italian Criticality Guide) and its validation test. (author)

15 CFR 1400.6 - Construction.

Science.gov (United States)

2010-01-01

... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Construction. 1400.6 Section 1400.6 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.6 Construction. Nothing in...

ARP1400 DVI break analysis using the MARS 3.1 multi-D component

International Nuclear Information System (INIS)

Hwang, Moon-Kyu; Lim, Hong-Sik; Lee, Seung-Wook; Bae, Sung-Won; Chung, Bub-Dong

2006-01-01

The current version of MARS 3.1 has a multi-D component intended to simulate an asymmetric multidimensional fluid behavior in a reactor core, downcomer or in a steam generator, in a more realistic manner. The feature is implemented in the 1-D module of the code. As opposed to the cross flow junction modeling, the multi-D component allows for a lateral momentum transfer as well as a sheer stress. Thus, a full three-Dimensional analysis capability is available as in the case of RELAP5-3D or CATHARE. In this study the multi-D component is applied to the hypothetical accident of a DVI (Direct Vessel Injection) break in the APR1400 plant, and the results are analyzed

Inhibition of Pseudomonas aeruginosa virulence: characterization of the AprA-AprI interface and species selectivity.

Science.gov (United States)

Bardoel, Bart W; van Kessel, Kok P M; van Strijp, Jos A G; Milder, Fin J

2012-01-20

Pseudomonas aeruginosa secretes the virulence factor alkaline protease (AprA) to enhance its survival. AprA cleaves one of the key microbial recognition molecules, monomeric flagellin, and thereby diminishes Toll-like receptor 5 activation. In addition, AprA degrades host proteins such as complement proteins and cytokines. P. aeruginosa encodes a highly potent inhibitor of alkaline protease (AprI) that is solely located in the periplasm where it is presumed to protect periplasmic proteins against secreted AprA. We set out to study the enzyme-inhibitor interactions in more detail in order to provide a basis for future drug development. Structural and mutational studies reveal that the conserved N-terminal residues of AprI occupy the protease active site and are essential for inhibitory activity. We constructed peptides mimicking the N-terminus of AprI; however, these were incapable of inhibiting AprA-mediated flagellin cleavage. Furthermore, we expressed and purified AprI of P. aeruginosa and the homologous (37% sequence identity) AprI of Pseudomonas syringae, which remarkably show species specificity for their cognate protease. Exchange of the first five N-terminal residues between AprI of P. syringae and P. aeruginosa did not affect the observed specificity, whereas exchange of only six residues located at the AprI surface that contacts the protease did abolish specificity. These findings are elementary steps toward the design of molecules derived from the natural inhibitor of the virulence factor AprA and their use in therapeutic applications in Pseudomonas and other Gram-negative infections. Copyright © 2011 Elsevier Ltd. All rights reserved.

21 CFR 184.1400 - Lecithin.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Lecithin. 184.1400 Section 184.1400 Food and Drugs... Substances Affirmed as GRAS § 184.1400 Lecithin. (a) Commercial lecithin is a naturally occurring mixture of... isolated as a gum following hydration of solvent-extracted soy, safflower, or corn oils. Lecithin is...

Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

International Nuclear Information System (INIS)

Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

2011-01-01

Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

Development of a computerized procedure system on the basis of formalized action and check

International Nuclear Information System (INIS)

Jung, Y. S.; Shin, Y. C.; Jung, K. H.; Sung, C. H.

1999-01-01

A computerized procedure system was developed on the basis of formalized action and check. The actions and checks are rendered into two dimensional flowchart with foldable, dynamic, and interactive functions. The action and check are evaluated by computer automatically, and can be overridden by operator if necessary. Detail of each action and check is described in the logic tree with several boolean instruction set and n out m of mathematical operator. Computer supports operators to execute the boolean instruction providing relevant process parameters and control devices. In spite of all these computer support, its inner logic is simple and transparent to operators to keep them in the control loop. The boolean instruction can have 3 states. And n out m operator is defined on the 3 states. The definition minimizes not only number of context switching in procedure execution, but also enables evaluation of the action and check. The process parameters are presented in both text label and graphical symbol to improve readability. Moreover, new mechanism to access any part of a procedure is devised while keeping the main stream of procedure execution

Computerized operating procedures for shearing and dissolution of segments from LWBR [Light Water Breeder Reactor] fuel rods

International Nuclear Information System (INIS)

Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

1987-05-01

This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work

7 CFR 1400.103 - Charitable organizations.

Science.gov (United States)

2010-01-01

... 7 Agriculture 10 2010-01-01 2010-01-01 false Charitable organizations. 1400.103 Section 1400.103... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.103 Charitable organizations. (a) A charitable organization, including a club, society, fraternal organization, or religious...

Probabilistics since WASH-1400

International Nuclear Information System (INIS)

Whitehead, N.E.

1980-01-01

Literature since the issuing of WASH-1400 reactor safety study shows that although the methodology has been attacked, it stands criticism well. Contrary to the aim of the study, which was to give a realistic, rather than a conservative risk estimate, there are many conservatisms in it. The strongly attacked treatment of common mode failure involving the square bounding model is shown here to be very likely to give correct results - and the applications of it in WASH-1400 do not often give results different from using the mean instead of the median. The Three-Mile Island accident is not such as to change the conclusions of WASH-1400 regarding core melt probabilities

21 CFR 886.1400 - Maddox lens.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Maddox lens. 886.1400 Section 886.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES OPHTHALMIC DEVICES Diagnostic Devices § 886.1400 Maddox lens. (a) Identification. A Maddox lens is a device...

7 CFR 1400.101 - Minor children.

Science.gov (United States)

2010-01-01

... 7 Agriculture 10 2010-01-01 2010-01-01 false Minor children. 1400.101 Section 1400.101 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.101 Minor children. (a) Except as provided in paragraph (b) of this section, payments received by a child under 18 years of age as of April 1...

Empirical investigation of workloads of operators in advanced control rooms

International Nuclear Information System (INIS)

Kim, Yochan; Jung, Wondea; Kim, Seunghwan

2014-01-01

This paper compares the workloads of operators in a computer-based control room of an advanced power reactor (APR 1400) nuclear power plant to investigate the effects from the changes in the interfaces in the control room. The cognitive-communicative-operative activity framework was employed to evaluate the workloads of the operator's roles during emergency operations. The related data were obtained by analyzing the tasks written in the procedures and observing the speech and behaviors of the reserved operators in a full-scope dynamic simulator for an APR 1400. The data were analyzed using an F-test and a Duncan test. It was found that the workloads of the shift supervisors (SSs) were larger than other operators and the operative activities of the SSs increased owing to the computer-based procedure. From these findings, methods to reduce the workloads of the SSs that arise from the computer-based procedure are discussed. (author)

X-ray computerized tomography based on the method of reciprocal projection with filtration by double differentiation. Procedure and information peculiarities

International Nuclear Information System (INIS)

Vajnberg, Eh.I.; Kazak, I.A.; Fajngojz, M.L.

1985-01-01

The results of experimental evaluation of procedure and information peculiarities of the method of reciprocal projection with filtration of projections by double differentiation (RPFDD) for the monitoring of industrial products are generalized. Succession of stages n the RPFDD method permits to separately optimize reconstruction of high-frequeny and low-frequency tomogram structure which sharply reduces the acuteness of contradictions of conventional scheme between the required increase of accuracy and intolerable growth of computerized charges. High accuracy of evaluation of the linear attenuation factor of low-frequency structures in a wide range of densities at the last stage of RPFDD scheme is attained despite the non-conventionally small range of values at earlier stages of computerized processing

29 CFR 1400.735-60 - Disciplinary actions.

Science.gov (United States)

2010-07-01

... 29 Labor 4 2010-07-01 2010-07-01 false Disciplinary actions. 1400.735-60 Section 1400.735-60 Labor..., RESPONSIBILITIES, AND DISCIPLINE Disciplinary Actions and Penalties § 1400.735-60 Disciplinary actions. The Service shall take prompt disciplinary action against an employee committing prohibited activity, or whose...

21 CFR 582.1400 - Lecithin.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Lecithin. 582.1400 Section 582.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS... Lecithin. (a) Product. Lecithin. (b) Conditions of use. This substance is generally recognized as safe when...

Hand-held electronic data collection and procedure environment

International Nuclear Information System (INIS)

Kennedy, E.; Doniz, K.

1996-01-01

As part of a CANDU Owners Group project, AECL has developed a hand-held electronic data collection and procedure environment. Integral to this environment is the C omputerized Procedure Engine . The development of the CPE allows operators, maintainers, and technical support staff to execute virtually any type of station procedure on a general-purpose PC-compatible hand-held computer. There are several advantages to using the computerized procedures: less paper use and handling, reduction in human error, reduction in rework in the field, an increase in procedural compliance, and immediate availability of data to download to databases and plant information systems. The paper describes: the advantages of using computerized procedures, why early forms of computerized procedures were inadequate, the features that the C omputerized Procedure Engine o ffers to the user, the streamlined life cycle of a computerized procedure, and field experience. The paper concludes that computerized procedures are ready for pilot applications at stations. (author)

Comparison of the effect of paper and computerized procedures on operator error rate and speed of performance

International Nuclear Information System (INIS)

Converse, S.A.; Perez, P.B.; Meyer, S.; Crabtree, W.

1994-01-01

The Computerized Procedures Manual (COPMA-II) is an advanced procedure manual that can be used to select and execute procedures, to monitor the state of plant parameters, and to help operators track their progress through plant procedures. COPMA-II was evaluated in a study that compared the speed and accuracy of operators' performance when they performed with COPMA-II and traditional paper procedures. Sixteen licensed reactor operators worked in teams of two to operate the Scales Pressurized Water Reactor Facility at North Carolina State University. Each team performed one change of power with each type of procedure to simulate performance under normal operating conditions. Teams then performed one accident scenario with COPMA-II and one with paper procedures. Error rates, performance times, and subjective estimates of workload were collected, and were evaluated for each combination of procedure type and scenario type. For the change of power task, accuracy and response time were not different for COPMA-II and paper procedures. Operators did initiate responses to both accident scenarios fastest with paper procedures. However, procedure type did not moderate response completion time for either accident scenario. For accuracy, performance with paper procedures resulted in twice as many errors as did performance with COPMA-II. Subjective measures of mental workload for the accident scenarios were not affected by procedure type

Computerized operation of the DIII-D neutral beams

International Nuclear Information System (INIS)

Glad, A.S.; Tooker, J.F.

1986-01-01

Operation of the DIII-D neutral beams utilizes computerized control to provide routine tokamak beam heating shots and an effective method for automatic ion source operation. Computerized control reduces operational complexity, thus providing consistent reliability and availability of beams and a significant reduction in the the costs of routine operation. The objectives in implementing computerized control for operation were: (1) to improve operator efficiency for controlling multiple beam lines and increasing beam availability through standard procedures, (2) to provide a simplified scheme that operators and coordinators can construct and maintain, and (3) to provide a single integrated mechanism for both tokamak operation and automatic source conditioning. The years of experience in operating neutral beams at Doublet III provided the data necessary to meet the objectives. The method for computerized control consisted of three integrated functions: (1) a structured command language was implemented to provide the mechanism for automatically sequencing beams, (2) a historical file was constructed from the operational parameters to characterize the ion source, and consists of data from approximately 100,000 beam shots, and (3) procedures were developed integrating the language to the historical file for normal operation and source conditioning. This paper describes the method for sequencing beams automatically, the structure of the historical data file, and the procedures which integrate the historical data with tokamak operation and automatic source conditioning

7 CFR 1400.104 - Changes in farming operations.

Science.gov (United States)

2010-01-01

... farming operation, (ii) The sale or gift of the equipment was based on the equipment's fair market value... 7 Agriculture 10 2010-01-01 2010-01-01 false Changes in farming operations. 1400.104 Section 1400... 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.104 Changes in farming...

7 CFR 1400.212 - Growers of hybrid seed.

Science.gov (United States)

2010-01-01

... 7 Agriculture 10 2010-01-01 2010-01-01 false Growers of hybrid seed. 1400.212 Section 1400.212... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.212 Growers of hybrid seed. The existence of a hybrid seed contract for a person or legal entity will not be taken into account...

21 CFR 876.1400 - Stomach pH electrode.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Stomach pH electrode. 876.1400 Section 876.1400...) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Diagnostic Devices § 876.1400 Stomach pH electrode. (a) Identification. A stomach pH electrode is a device used to measure intragastric and intraesophageal pH (hydrogen...

30 CFR 77.1400 - Personnel hoists and elevators.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Personnel hoists and elevators. 77.1400 Section 77.1400 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS, SURFACE COAL MINES AND SURFACE WORK AREAS OF UNDERGROUND COAL MINES Personnel Hoisting § 77.1400 Personnel...

Evaluation of Ablation rate by the change of Sacrificial Material for PECS in EU-APR

International Nuclear Information System (INIS)

Hwang, Do Hyun; Kim, Yong Soo; Lee, Keun Sung

2015-01-01

EU-APR, modified and improved from its original design of APR1400, has been developed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. In EU-APR, Severe Accident Mitigation Systems are dedicated to providing an independent defense line from that of Engineered Safety Feature (ESF) and Diverse Safety Feature (DSF). They consist of Emergency Reactor Depressurization System (ERDS), Passive Ex-vessel corium retaining and Cooling System (PECS), Severe Accident Containment Spray System (SACSS), Hydrogen Mitigation System (HMS) and Containment Filtered Vent System (CFVS). The PECS, so called core catcher, was introduced to prevent the Molten Core Concrete Interaction (MCCI) after Reactor Vessel (RV) failure. The PECS has experienced a lot of changes from its original design. Recently, the most significant change was that as a SM, limestone concrete is installed on PECS's body wall instead of previous sacrificial material rich in Fe 2 O 3 . The main reason of this design change is to overcome the issue that the sacrificial material is ablated rather too fast when reacting with corium that contains a large fraction of Zr metal. Other changes in the geometry of PECS's wall and downcomer design are considered as minor ones. In this paper, the comparison of ablation rates between previous SM and limestone concrete is carried out using MAAP5 code with respective MCCI model according to the material. In this paper, major improvements of MAAP5 model for PECS in EU-APR are presented and the evaluation of ablation rate for the previous SM model and the new LC model is carried out by means of ablation depths with LBLOCA sequence. Two models have respective unique ablation process. The ablation of LC model proceeds at a constant rate regardless of water while the ablation of SM model proceeds at a faster rate before the arrival of cooling water for corium and SM mixture. The change of sacrificial material also

Computerized management of plant intervention tasks

International Nuclear Information System (INIS)

Remacle, J.; Quoidbach, G.

1993-01-01

The concept of 'computerized management' of plant intervention tasks was developed by TRACTEBEL in 1983 for the Belgian power plants of ELECTRABEL. The main objective of the 'Computerized Management of Plant Intervention Tasks' is to help the staff of a nuclear or a conventional power plant in planning, organizing, and carrying out any (preventive or corrective) maintenance task. It consists of a group of interconnected functional modules acting on a unique and homogeneous data base. A short description of 3 modules is given, i.e., the 'User' Module, the 'Equipment' Module and the 'Periodic Procedure' Module. (Z.S.)

48 CFR 22.1400 - Scope of subpart.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Scope of subpart. 22.1400 Section 22.1400 Federal Acquisition Regulations System FEDERAL ACQUISITION REGULATION SOCIOECONOMIC PROGRAMS APPLICATION OF LABOR LAWS TO GOVERNMENT ACQUISITIONS Employment of Workers with Disabilities 22...

Computerized tomography

International Nuclear Information System (INIS)

Rubashov, I.B.

1985-01-01

Operating principle is described for the devices of computerized tomography used in medicine for diagnosis of brain diseases. Computerized tomography is considered as a part of computerized diagnosis, as a part of information science. It is shown that computerized tomography is a real existed field of investigations in medicine and industrial production

29 CFR 1400.735-3 - Advice and counseling service.

Science.gov (United States)

2010-07-01

... 29 Labor 4 2010-07-01 2010-07-01 false Advice and counseling service. 1400.735-3 Section 1400.735-3 Labor Regulations Relating to Labor (Continued) FEDERAL MEDIATION AND CONCILIATION SERVICE STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE General § 1400.735-3 Advice and counseling service...

12 CFR 1400.1 - Farm Credit System Insurance Corporation.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 7 2010-01-01 2010-01-01 false Farm Credit System Insurance Corporation. 1400.1 Section 1400.1 Banks and Banking FARM CREDIT SYSTEM INSURANCE CORPORATION ORGANIZATION AND FUNCTIONS Organization and Functions § 1400.1 Farm Credit System Insurance Corporation. The Farm Credit...

Computerizing Maintenance Management Improves School Processes.

Science.gov (United States)

Conroy, Pat

2002-01-01

Describes how a Computerized Maintenance Management System (CMMS), a centralized maintenance operations database that facilitates work order procedures and staff directives, can help individual school campuses and school districts to manage maintenance. Presents the benefits of CMMS and things to consider in CMMS selection. (EV)

Integrated System Validation of Barakah Nuclear Power Plant in UAE for The Human Factor Engineering

Energy Technology Data Exchange (ETDEWEB)

Kim, Munsoo [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

APR1400 simulator has been developed based on the state-of-the-art object-oriented simulation technology of TH(Thermo-Hydraulic) and Reactor Core model, which is applied for the first time in the our country and for the exportation, to well simulate characteristics of APR1400. Barakah unit 1,2 simulator are constructed and supplied with this type simulator model. Integrated system validation was performed using a simulator to verify the HFE(Human Factor Engineering) design of the MCR(Maim Control Room) for instrumentation and control system validation of the UAE nuclear power plant. APR1400 for the Barakah unit 1,2 has many specific features such as digital I and C, and digitalized main control room (MCR) design. From January 2016 to February, during six weeks, the tests carried out three times repeatedly and the various proposals for ergonomical satisfactation were derived. However, the HFE errors that cause significant change of validation target for APR1400 MCR design safety fidelity wasn't found. This has resulted in the conclusion to prove the stability of the basic design of APR1400 MCR. In the future, using the simulator derives the HFE requirements of the MCR systems and continually improve the simulator will be built in close to real high-fidelity power plant. These Integrated system validations are likely to be a great help in operating safety and preventing human errors by operators. Therefore successful completion of the Integrated System Validation for BNPP simulation will be effective to promotion the distinction of our simulator and APR1400 NPP.

29 CFR 1400.735-19 - Influencing Members of Congress.

Science.gov (United States)

2010-07-01

... 29 Labor 4 2010-07-01 2010-07-01 false Influencing Members of Congress. 1400.735-19 Section 1400... Responsibilities § 1400.735-19 Influencing Members of Congress. No money appropriated to the Service shall be used... devices intended to influence any Member of Congress regarding any legislation or appropriation before the...

Development of core technology for KNGR system design

International Nuclear Information System (INIS)

Lee, K. Y; Park, J. K.; Ham, C. S.

2002-05-01

For evaluating the design issues in the advanced MMI, we have studied the selection of evaluation issues, evaluation schedules, and the application method of evaluation results from the beginning of APR1400 MMI design project. For producing input for human reliability analysis, we proposed a new error taxonomy and approach to explain the new human error potential. In order to evaluate whether or not the lifetime of APR1400 I and C system meet 60 years, the lifetime of the system has been calculated by using failure rates in this research. In order to qualify the software of APR1400 protection system, the detail scope and contents are to develop the methods, to evaluate the design results using the methods, and to establish a database for the evaluation. In external vessel cooling strategy to maintain the reactor vessel integrity in a severe accident of the APR 1400, an evaluation methodology was developed, and the failure possibility of penetrations and failure mode of the reactor vessel penetrations have been estimated through experiments and analysis. Two types of experiments have been performed using alumina melt as a simulant and a sustained heating method using an inductor, and their results have been evaluated along with APR 1400 using the LILAC, the FLUENT, and the ABAQUS computer codes

Assistant Personal Robot (APR: Conception and Application of a Tele-Operated Assisted Living Robot

Directory of Open Access Journals (Sweden)

Eduard Clotet

2016-04-01

Full Text Available This paper presents the technical description, mechanical design, electronic components, software implementation and possible applications of a tele-operated mobile robot designed as an assisted living tool. This robotic concept has been named Assistant Personal Robot (or APR for short and has been designed as a remotely telecontrolled robotic platform built to provide social and assistive services to elderly people and those with impaired mobility. The APR features a fast high-mobility motion system adapted for tele-operation in plain indoor areas, which incorporates a high-priority collision avoidance procedure. This paper presents the mechanical architecture, electrical fundaments and software implementation required in order to develop the main functionalities of an assistive robot. The APR uses a tablet in order to implement the basic peer-to-peer videoconference and tele-operation control combined with a tactile graphic user interface. The paper also presents the development of some applications proposed in the framework of an assisted living robot.

40 CFR 1400.13 - Read-only database.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Read-only database. 1400.13 Section... INFORMATION Other Provisions § 1400.13 Read-only database. The Administrator is authorized to establish... public off-site consequence analysis information by means of a central database under the control of the...

Computerized tomography

International Nuclear Information System (INIS)

1980-01-01

Improvements in the design of computerized tomographic X-ray equipment are described which lead to improvements in the mechanical properties, speed and size of scanning areas. The method envisages the body being scanned as a two-dimensional matrix of elements arising from a plurality of concentric rings. The concentric centre need not coincide with the axis of rotation. The procedures for rotation of the X-ray beam and detectors around the patient and for translating the measured information into attenuation coefficients for each matrix element of the body are described in detail. Explicit derivations are given for the mathematical formulae used. (U.K.)

Evaluation of Ablation rate by the change of Sacrificial Material for PECS in EU-APR

Energy Technology Data Exchange (ETDEWEB)

Hwang, Do Hyun; Kim, Yong Soo; Lee, Keun Sung [KHNP-CRI, Daejeon (Korea, Republic of)

2015-05-15

EU-APR, modified and improved from its original design of APR1400, has been developed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. In EU-APR, Severe Accident Mitigation Systems are dedicated to providing an independent defense line from that of Engineered Safety Feature (ESF) and Diverse Safety Feature (DSF). They consist of Emergency Reactor Depressurization System (ERDS), Passive Ex-vessel corium retaining and Cooling System (PECS), Severe Accident Containment Spray System (SACSS), Hydrogen Mitigation System (HMS) and Containment Filtered Vent System (CFVS). The PECS, so called core catcher, was introduced to prevent the Molten Core Concrete Interaction (MCCI) after Reactor Vessel (RV) failure. The PECS has experienced a lot of changes from its original design. Recently, the most significant change was that as a SM, limestone concrete is installed on PECS's body wall instead of previous sacrificial material rich in Fe{sub 2}O{sub 3}. The main reason of this design change is to overcome the issue that the sacrificial material is ablated rather too fast when reacting with corium that contains a large fraction of Zr metal. Other changes in the geometry of PECS's wall and downcomer design are considered as minor ones. In this paper, the comparison of ablation rates between previous SM and limestone concrete is carried out using MAAP5 code with respective MCCI model according to the material. In this paper, major improvements of MAAP5 model for PECS in EU-APR are presented and the evaluation of ablation rate for the previous SM model and the new LC model is carried out by means of ablation depths with LBLOCA sequence. Two models have respective unique ablation process. The ablation of LC model proceeds at a constant rate regardless of water while the ablation of SM model proceeds at a faster rate before the arrival of cooling water for corium and SM mixture. The change of sacrificial material

15 CFR 1400.4 - Evidence of social or economic disadvantage.

Science.gov (United States)

2010-01-01

... disadvantage. 1400.4 Section 1400.4 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign... ASSISTANCE § 1400.4 Evidence of social or economic disadvantage. (a) The representatives of the group requesting formal designation should establish social or economic disadvantage by a preponderance of the...

Pre-Test Analysis of Major Scenarios for ATLAS

Energy Technology Data Exchange (ETDEWEB)

Euh, Dong-Jin; Choi, Ki-Yong; Park, Hyun-Sik; Kwon, Tae-Soon

2007-02-15

A thermal-hydraulic integral effect test facility, ATLAS was constructed at the Korea Atomic Energy Research Institute (KAERI). The ATLAS is a 1/2 reduced height and 1/288 volume scaled test facility based on the design features of the APR1400. The simulation capability of the ATLAS for major design basis accidents (DBAs), including a large-break loss-of-coolant (LBLOCA), DVI line break and main steam line break (MSLB) accidents, is evaluated by the best-estimate system code, MARS, with the same control logics, transient scenarios and nodalization scheme. The validity of the applied scaling law and the thermal-hydraulic similarity between the ATLAS and the APR1400 for the major design basis accidents are assessed. It is confirmed that the ATLAS has a capability of maintaining an overall similarity with the reference plant APR1400 for the major design basis accidents considered in the present study. However, depending on the accident scenarios, there are some inconsistencies in certain thermal hydraulic parameters. It is found that the inconsistencies are mainly due to the reduced power effect and the increased stored energy in the structure. The present similarity analysis was successful in obtaining a greater insight into the unique design features of the ATLAS and would be used for developing the optimized experimental procedures and control logics.

Pre-Test Analysis of Major Scenarios for ATLAS

International Nuclear Information System (INIS)

Euh, Dong-Jin; Choi, Ki-Yong; Park, Hyun-Sik; Kwon, Tae-Soon

2007-02-01

A thermal-hydraulic integral effect test facility, ATLAS was constructed at the Korea Atomic Energy Research Institute (KAERI). The ATLAS is a 1/2 reduced height and 1/288 volume scaled test facility based on the design features of the APR1400. The simulation capability of the ATLAS for major design basis accidents (DBAs), including a large-break loss-of-coolant (LBLOCA), DVI line break and main steam line break (MSLB) accidents, is evaluated by the best-estimate system code, MARS, with the same control logics, transient scenarios and nodalization scheme. The validity of the applied scaling law and the thermal-hydraulic similarity between the ATLAS and the APR1400 for the major design basis accidents are assessed. It is confirmed that the ATLAS has a capability of maintaining an overall similarity with the reference plant APR1400 for the major design basis accidents considered in the present study. However, depending on the accident scenarios, there are some inconsistencies in certain thermal hydraulic parameters. It is found that the inconsistencies are mainly due to the reduced power effect and the increased stored energy in the structure. The present similarity analysis was successful in obtaining a greater insight into the unique design features of the ATLAS and would be used for developing the optimized experimental procedures and control logics

Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

International Nuclear Information System (INIS)

Hwang, J. S.; Suh, K. Y.

2009-01-01

The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

Pre-test analysis of a LBLOCA using the design data of the ATLAS facility, a reduced-height integral effect test loop for PWRs

International Nuclear Information System (INIS)

Hyun-Sik Park; Ki-Yong Choi; Dong-Jin Euh; Tae-Soon Kwon; Won-Pil Baek

2005-01-01

Full text of publication follows: The simulation capability of the KAERI integral effect test facility, ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation), has been assessed for a large-break loss-of-coolant accident (LBLOCA) transient. The ATLAS facility is a 1/2 height-scaled, 1/144 area-scaled (1/288 in volume scale), and full-pressure test loop based on the design features of the APR1400, an evolutionary pressurized water reactor that has been developed by Korean industry. The APR1400 has four mechanically separated hydraulic trains for the emergency core cooling system (ECCS) with direct vessel injection (DVI). The APR1400 design features have brought about several new safety issues related to the LBLOCA including the steam-water interaction, ECC bypass, and boiling in the reactor vessel downcomer. The ATLAS facility will be used to investigate the multiple responses between the systems or between the components during various anticipated transients. The ATLAS facility has been designed according to a scaling method that is mainly based on the model suggested by Ishii and Kataoka. The ATLAS facility is being evaluated against the prototype plant APR1400 with the same control logics and accident scenarios using the best-estimated code, MARS. This paper briefly introduces the basic design features of the ATLAS facility and presents the results of pre-test analysis for a postulated LBLOCA of a cold leg. The LBLOCA analyses has been conducted to assess the validity of the applied scaling law and the similarity between the ATLAS facility and the APR1400. As the core simulator of the ATLAS facility has the 10% capability of the scaled full power, the blowdown phase can not be simulated, and the starting point of the accident scenario is around the end of blowdown. So it is an important problem to find the correct initial conditions. For the analyzed LBLOCA scenario, the ATLAS facility showed very similar thermal-hydraulic characteristics to the APR

Guidelines for the presentation of emergency operating procedures using advanced information technology

International Nuclear Information System (INIS)

Green, M.; Hollnagel, E.; Niwa, Yuji

1994-01-01

New methods of information presentation and interface design are changing the conditions for work in the modern NPP control room. One area receiving considerable attention is that of Emergency Operating Procedures (EOP). Conventionally such procedures are presented using hard copy manuals; however, development in IT means that there are new opportunities for the computerization of such procedures. This paper reports on the development of human factors guidelines for the computerized presentation of EOPs. After identifying the principle stages in the transition from procedures as documents to fully automated procedures, computerised procedure presentation is briefly discussed. Guidelines for the presentation of such procedures are outlined starting with the high level goals for guidelines themselves. Such goals also constitute the criteria against which the computerized procedures are measured during implementation. Six dimensions describing computerized procedure presentation are presented and two are explore in more detail by identifying points along each dimension that characterize different levels of IT sophistication. (author). 8 refs, 1 tab

The impact of WASH-1400 on reactor safety evaluation

International Nuclear Information System (INIS)

Tanguy, P.Y.

1976-01-01

Trends in reactor safety evaluation in France following the publication of WASH-1400 (the Rasmussen Report) are presented. What is called 'the meteorite case' is first schematically presented as follows: WASH-1400 shows nuclear risk equivalent to meteorite risk and reasonable corrections cannot make many orders of magnitude, consequently present safety rules are adequate. The very impact of WASH-1400 on safety approach is then discussed as for: assistance to deterministic safety analysis, introduction of probabilistic safety criteria, acceptable level of risk, and the use of results in research and reactor operating experience

Antenna Pattern Range (APR)

Data.gov (United States)

Federal Laboratory Consortium — TheAntenna Pattern Range (APR)features a non-metallic arch with a trolley to move the transmit antenna from the horizon to zenith. At the center of the ground plane,...

A Computerized Approach to Trickle-Process, Random Assignment.

Science.gov (United States)

Braucht, G. Nicholas; Reichardt, Charles S.

1993-01-01

Procedures for implementing random assignment with trickle processing and ways they can be corrupted are described. A computerized method for implementing random assignment with trickle processing is presented as a desirable alternative in many situations and a way of protecting against threats to assignment validity. (SLD)

Preliminary ripple effect analysis for HTR 350MWt 4 modules construction

Energy Technology Data Exchange (ETDEWEB)

Lee, T. H.; Lee, K. Y.; Shin, Y. J. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

We propose quantitative analysis techniques for ripple effects such as the production inducement effect and employment inducement effect for HTR 350MWt x 4 module construction and operation ripple effect based on NOAK. It is known that APR1400 reactors export ripple effect is about 8,500 billion KRW. As a result, HTR construction has more effective effect than that of APR1400.

Experience feedback of computerized controlled nuclear power plants

International Nuclear Information System (INIS)

Poizat, F.

2004-01-01

The N4 step of French PWR-type nuclear power plants is characterized by an instrumentation and control system entirely computerized (operation procedures including normal and accidental operation). Four power plants of this type (Chooz and Civaux sites) of 1450 MWe each were connected to the power grid between August 1996 and December 1999. The achievement of this program make it possible and necessary to carry out an experience feedback about the development, successes and difficulties encountered in order to draw out some lessons for future realizations. This is the aim of this article: 1 - usefulness and difficulties of such an experience feedback: evolution of instrumentation and control systems, necessary cautions; 2 - a successful computerized control: checking of systems operation, advantages, expectations; 3 - efficiency of computerized systems: demonstration of operation safety, profitability; 4 - conclusions and interrogations: system approach instead of 'micro-software' approach, commercial or 'made to measure' products, contract agreement with a supplier, when and how upgrading. (J.S.)

Evaluation of Thermal Load to the Lower Head Vessel Using the ASTEC Computer Code

International Nuclear Information System (INIS)

Park, Raejoon; Ahn, Kwangil

2013-01-01

The thermal load from the corium to the lower head vessel in the APR (Advanced Power reactor) 1400 during a small break loss of coolant accident (SBLOCA) without a safety injection (SI) has been evaluated using the ASTEC (Accident Source Term Evaluation Code) computer code, which has been developed as a part of the EU (European Union)-SARNET (Severe Accident Research NET work) program. The ASTEC results predict that the reactor vessel did not fail by using an ERVC, in spite of the large melting of the reactor vessel wall in a two-layer formation case of the SBLOCA in the APR1400. The outer surface conditions of the temperature and heat transfer coefficient are not effective on the vessel geometry change, which are preliminary results. A more detailed analysis of the main parameter effects on the corium behavior in the lower plenum is necessary to evaluate the IVR-ERVC in the APR1400, in particular, for a three-layer formation of the TLFW. Comparisons of the present results with others are necessary to verify and apply them to the actual IVR-ERVC evaluation in the APR1400

Pseudomonas syringae evades host immunity by degrading flagellin monomers with alkaline protease AprA.

Science.gov (United States)

Pel, Michiel J C; van Dijken, Anja J H; Bardoel, Bart W; Seidl, Michael F; van der Ent, Sjoerd; van Strijp, Jos A G; Pieterse, Corné M J

2014-07-01

Bacterial flagellin molecules are strong inducers of innate immune responses in both mammals and plants. The opportunistic pathogen Pseudomonas aeruginosa secretes an alkaline protease called AprA that degrades flagellin monomers. Here, we show that AprA is widespread among a wide variety of bacterial species. In addition, we investigated the role of AprA in virulence of the bacterial plant pathogen P. syringae pv. tomato DC3000. The AprA-deficient DC3000 ΔaprA knockout mutant was significantly less virulent on both tomato and Arabidopsis thaliana. Moreover, infiltration of A. thaliana Col-0 leaves with DC3000 ΔaprA evoked a significantly higher level of expression of the defense-related genes FRK1 and PR-1 than did wild-type DC3000. In the flagellin receptor mutant fls2, pathogen virulence and defense-related gene activation did not differ between DC3000 and DC3000 ΔaprA. Together, these results suggest that AprA of DC3000 is important for evasion of recognition by the FLS2 receptor, allowing wild-type DC3000 to be more virulent on its host plant than AprA-deficient DC3000 ΔaprA. To provide further evidence for the role of DC3000 AprA in host immune evasion, we overexpressed the AprA inhibitory peptide AprI of DC3000 in A. thaliana to counteract the immune evasive capacity of DC3000 AprA. Ectopic expression of aprI in A. thaliana resulted in an enhanced level of resistance against wild-type DC3000, while the already elevated level of resistance against DC3000 ΔaprA remained unchanged. Together, these results indicate that evasion of host immunity by the alkaline protease AprA is important for full virulence of strain DC3000 and likely acts by preventing flagellin monomers from being recognized by its cognate immune receptor.

Applicability of the Reactor Safety Study (WASH-1400) to LMFBR risk assessments

International Nuclear Information System (INIS)

El-Sheikh, K.A.; Feller, K.G.; Fleischer, L.; Greebler, P.; McDonald, A.; Sultan, P.; Temme, M.I.; Fullwood, R.R.

1976-01-01

The feasibility of applying the WASH-1400 methods and data to LMFBR risk assessment is evaluated using the following approach for a selected LMFBR: (1) Structuring the LMFBR risk assessment problem in a modular form similar to WASH-1400; (2) Comparing the predictive tools applicable to each module; (3) Comparing the dependencies among the various modules. It is concluded that the WASH-1400 applicability is limited due to LWR-LMFBR differences in operating environments and accident phenomena. WASH-1400 and LMFBR specific methods applicable to LMFBR risk assessments are indicated

EU-APR Design in compliance with EUR Grid Requirement

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Hwan; Lee, Keun-Sung [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

European Utility Requirements (EUR) provides technical requirements for the generation III nuclear power plant in the European countries. EUR grid requirements present the plant requirements to satisfy the needs of the grid network. The grid requirements are the precondition for the operation of a generating plant on the network. This paper describes EU-APR design which has taken account of EUR grid requirements. In this paper, EU-APR designs according to the EUR grid requirements were described. EU-APR was designed in compliance with the voltage and frequency operation field and also designed to have the capability of load following such as primary control, secondary control, and daily load following. Consequently, the EU-APR design according to the EUR grid requirements is expected to get competitiveness and enhance the license feasibility in the European nuclear market.

Contribution of computerization to alarm processing: A French safety view

Energy Technology Data Exchange (ETDEWEB)

Cette, W [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

1997-09-01

Following the TMI accident and according to the requirement of the French safety authority, very important studies were performed by the French utility, Electricite de France (EDF), and assessed by the Institute for Nuclear Safety and Protection (IPSN) on reactor operation in conventional control rooms, particularly on alarm processing. These studies dealt with the man-machine interface, as well as design and exploitation requirements, presentation and management of alarm signals, and associated operating documents. The conclusions of these studies have led to improvements in French conventional control rooms. The current state of these control rooms and links between alarm sets and operating documents will be shortly presented in the first part of the paper. More recently, the computerized means implemented in the PWR 1400 MWe control rooms (N4) profoundly modified reactor operation. In particular, major advances concern alarm processing in comparison with conventional control rooms. The N4 plants provide a more rigorous approach in processing and presentation of alarms than in the past. Indeed, EDF wanted to have less alarms switched on during plant upsets and to make them more characteristic of a specific situation of the process. For example, computerization makes it easier to validate or inhibit alarms according to the situation, to allow the operator to manage alarm presentation and to propose on-line alarm sheets to the operator etc. This approach in comparison with conventional control rooms, and the IPSN assessment will be presented in the second part of this paper. (author).

Contribution of computerization to alarm processing: A French safety view

International Nuclear Information System (INIS)

Cette, W.

1997-01-01

Following the TMI accident and according to the requirement of the French safety authority, very important studies were performed by the French utility, Electricite de France (EDF), and assessed by the Institute for Nuclear Safety and Protection (IPSN) on reactor operation in conventional control rooms, particularly on alarm processing. These studies dealt with the man-machine interface, as well as design and exploitation requirements, presentation and management of alarm signals, and associated operating documents. The conclusions of these studies have led to improvements in French conventional control rooms. The current state of these control rooms and links between alarm sets and operating documents will be shortly presented in the first part of the paper. More recently, the computerized means implemented in the PWR 1400 MWe control rooms (N4) profoundly modified reactor operation. In particular, major advances concern alarm processing in comparison with conventional control rooms. The N4 plants provide a more rigorous approach in processing and presentation of alarms than in the past. Indeed, EDF wanted to have less alarms switched on during plant upsets and to make them more characteristic of a specific situation of the process. For example, computerization makes it easier to validate or inhibit alarms according to the situation, to allow the operator to manage alarm presentation and to propose on-line alarm sheets to the operator etc. This approach in comparison with conventional control rooms, and the IPSN assessment will be presented in the second part of this paper. (author)

Design of a Load Following Controller for APR+ Nuclear Plants

International Nuclear Information System (INIS)

Lee, Sim Won; Kim, Jae Hwan; Kim, Dong Su; Na, Man Gyun; Yu, Keuk Jong

2011-01-01

The load-following operation of an APR+ nuclear plants is required to restrain the adjustment of boric acid concentration and to efficiently control the control rods for the flexibility of the operation. Especially, axial flux distribution disproportion that is usually caused by load-following operation in a reactor core induces xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. Rapid and smooth power maneuvering has its benefits in view of the economical and safe operation of reactors, so it is required that the controller is efficiently designed. Therefore, the load-following operation of an APR+ nuclear plants needs the ultimate automatic control and the advanced control method that satisfies the conditions such as the flexibility, safety and convenience. A model predictive control (MPC) method is applied to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for an APR+ nuclear plants. Some tracking controllers use only the current tracking command. On the other hand, since MPC considers future commands in addition to the current tracking command, the MPC can achieve better tracking performance. Therefore, the MPC has been applied very much to the control of industrial process systems. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the second control input is not implemented and the procedure to solve the optimization problem is then repeated. The power level and the ASI are controlled by the regulating control banks and part-strength control banks together with the automatic adjustment of boric acid concentration. The 3-dimensional MASTER code, which models the APR+ nuclear

Design of a Load Following Controller for APR+ Nuclear Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Sim Won; Kim, Jae Hwan; Kim, Dong Su; Na, Man Gyun [Chosun University, Gwnagju (Korea, Republic of); Yu, Keuk Jong [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2011-08-15

The load-following operation of an APR+ nuclear plants is required to restrain the adjustment of boric acid concentration and to efficiently control the control rods for the flexibility of the operation. Especially, axial flux distribution disproportion that is usually caused by load-following operation in a reactor core induces xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. Rapid and smooth power maneuvering has its benefits in view of the economical and safe operation of reactors, so it is required that the controller is efficiently designed. Therefore, the load-following operation of an APR+ nuclear plants needs the ultimate automatic control and the advanced control method that satisfies the conditions such as the flexibility, safety and convenience. A model predictive control (MPC) method is applied to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for an APR+ nuclear plants. Some tracking controllers use only the current tracking command. On the other hand, since MPC considers future commands in addition to the current tracking command, the MPC can achieve better tracking performance. Therefore, the MPC has been applied very much to the control of industrial process systems. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the second control input is not implemented and the procedure to solve the optimization problem is then repeated. The power level and the ASI are controlled by the regulating control banks and part-strength control banks together with the automatic adjustment of boric acid concentration. The 3-dimensional MASTER code, which models the APR+ nuclear

A Probability Analysis of the Generating Cost for APR1000+

Energy Technology Data Exchange (ETDEWEB)

Ha, Gag-Hyeon; Kim, Dae-Hun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, cutting CO{sub 2} emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the APR1000+ development is to ensure export competitiveness in the developing countries in the Middle East and Southeast Asia. To that end, APR1000+(1,000MWe, 3.5 generation) will be developed based on APR+ (1,500MWe, 3.5 generation). And comparing to OPR1000(Korean Standard Nuclear power Plant, 2.5 generation), APR1000+ have many design features such as the 60 year design life time, comprehensive site requirement of 0.3g seismic design, stability improvement, operability improvement and provisions for severe accidents. In this simulation, the results of generating cost for APR1000+ preliminary conceptual design using a probability method was shown to be 48.37 ~ 74.22 won/kWh(median value 56.51 won/kWh). Those of OPR1000 was 42.08 ~ 61.77 won/kWh(median value 48.63 won/kWh). APR1000+ has -16.2% cost advantage over OPR1000 nuclear power plant. The main reason of this results is due to adding several safety designs.

The trauma registry compared to All Patient Refined Diagnosis Groups (APR-DRG).

Science.gov (United States)

Hackworth, Jodi; Askegard-Giesmann, Johanna; Rouse, Thomas; Benneyworth, Brian

2017-05-01

Literature has shown there are significant differences between administrative databases and clinical registry data. Our objective was to compare the identification of trauma patients using All Patient Refined Diagnosis Related Groups (APR-DRG) as compared to the Trauma Registry and estimate the effects of those discrepancies on utilization. Admitted pediatric patients from 1/2012-12/2013 were abstracted from the trauma registry. The patients were linked to corresponding administrative data using the Pediatric Health Information System database at a single children's hospital. APR-DRGs referencing trauma were used to identify trauma patients. We compared variables related to utilization and diagnosis to determine the level of agreement between the two datasets. There were 1942 trauma registry patients and 980 administrative records identified with trauma-specific APR-DRG during the study period. Forty-two percent (816/1942) of registry records had an associated trauma-specific APR-DRG; 69% of registry patients requiring ICU care had trauma APR-DRGs; 73% of registry patients with head injuries had trauma APR-DRGs. Only 21% of registry patients requiring surgical management had associated trauma APR-DRGs, and 12.5% of simple fractures had associated trauma APR-DRGs. APR-DRGs appeared to only capture a fraction of the entire trauma population and it tends to be the more severely ill patients. As a result, the administrative data was not able to accurately answer hospital or operating room utilization as well as specific information on diagnosis categories regarding trauma patients. APR-DRG administrative data should not be used as the only data source for evaluating the needs of a trauma program. Copyright © 2016 Elsevier Ltd. All rights reserved.

WASH-1400: quantifying the uncertainties

International Nuclear Information System (INIS)

Erdmann, R.C.; Leverenz, F.L. Jr.; Lellouche, G.S.

1981-01-01

The purpose of this paper is to focus on the limitations of the WASH-1400 analysis in estimating the risk from light water reactors (LWRs). This assessment attempts to modify the quantification of the uncertainty in and estimate of risk as presented by the RSS (reactor safety study). 8 refs

7 CFR 1400.402 - Notification.

Science.gov (United States)

2010-01-01

... social security number or taxpayer identification number of such a person or legal entity, if known, and... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Foreign Persons § 1400.402 Notification. (a) Any legal entity... legal entity conducts its farming operation if: (1) Any person, group of persons, legal entity, or group...

Computerized adaptive testing item selection in computerized adaptive learning systems

NARCIS (Netherlands)

Eggen, Theodorus Johannes Hendrikus Maria; Eggen, T.J.H.M.; Veldkamp, B.P.

2012-01-01

Item selection methods traditionally developed for computerized adaptive testing (CAT) are explored for their usefulness in item-based computerized adaptive learning (CAL) systems. While in CAT Fisher information-based selection is optimal, for recovering learning populations in CAL systems item

45 CFR 310.40 - What requirements apply for accessing systems and records for monitoring Computerized Tribal IV-D...

Science.gov (United States)

2010-10-01

... 45 Public Welfare 2 2010-10-01 2010-10-01 false What requirements apply for accessing systems and records for monitoring Computerized Tribal IV-D Systems and Office Automation? 310.40 Section 310.40... COMPUTERIZED TRIBAL IV-D SYSTEMS AND OFFICE AUTOMATION Accountability and Monitoring Procedures for...

April 2006. 32 Major Orthopaedic Procedures

African Journals Online (AJOL)

user

2006-04-01

Apr 1, 2006 ... Major Orthopaedic Procedures: 17 Year Trends. Biruk Lambisso Wamisho1 ... financial and logistic constraints with poor compliance of ... Modern orthopaedic surgery is very expensive. A highly ..... Case management. Tribury.

Cognitive task analysis and the design of computerized operator aids

International Nuclear Information System (INIS)

Andersson, H.

1985-01-01

The new technological possibilities have led to the initiation of many projects for the design and evaluation of computerized operator support systems to be implemented in nuclear power plant control rooms. A typical finding so far has been that operators often have a positive attitude towards such systems but still don't use them very much, mostly because they find almost the same information on the conventional control boards which they are accustomed to use. Still, however, there is a widely shared belief that conventional control rooms have short-comings that make the use of computerized aids necessary. One reason for the limited success so far is that the new systems often are poorly integrated with the existing conventional instrumentation and with the working procedures. The reluctance to use new computer based aids, despite their nice features, is therefore probably caused by an inadequate task analysis made prior to the design of these computerized operator support systems

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

International Nuclear Information System (INIS)

Ha, Huiun; Suh, Jungsoo

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Ha, Huiun; Suh, Jungsoo [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

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

SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

2012-10-15

APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

Design-related influencing factors of the computerized procedure system for inclusion into human reliability analysis of the advanced control room

International Nuclear Information System (INIS)

Kim, Jaewhan; Lee, Seung Jun; Jang, Seung Cheol; Ahn, Kwang-Il; Shin, Yeong Cheol

2013-01-01

This paper presents major design factors of the computerized procedure system (CPS) by task characteristics/requirements, with individual relative weight evaluated by the analytic hierarchy process (AHP) technique, for inclusion into human reliability analysis (HRA) of the advanced control rooms. Task characteristics/requirements of an individual procedural step are classified into four categories according to the dynamic characteristics of an emergency situation: (1) a single-static step, (2) a single-dynamic and single-checking step, (3) a single-dynamic and continuous-monitoring step, and (4) a multiple-dynamic and continuous-monitoring step. According to the importance ranking evaluation by the AHP technique, ‘clearness of the instruction for taking action’, ‘clearness of the instruction and its structure for rule interpretation’, and ‘adequate provision of requisite information’ were rated as of being higher importance for all the task classifications. Importance of ‘adequacy of the monitoring function’ and ‘adequacy of representation of the dynamic link or relationship between procedural steps’ is dependent upon task characteristics. The result of the present study gives a valuable insight on which design factors of the CPS should be incorporated, with relative importance or weight between design factors, into HRA of the advanced control rooms. (author)

Pseudomonas syringae evades host Immunity by degrading flagellin monomers with alkaline protease AprA

OpenAIRE

Pel, M.J.C.; Van Dijken, A.J.H.; Bardoel, B.W.; Seidl, M.F; Van der Ent, S.; Van Strijp, J.A.G.

2014-01-01

Bacterial flagellin molecules are strong inducers of innate immune responses in both mammals and plants. The opportunistic pathogen Pseudomonas aeruginosa secretes an alkaline protease called AprA that degrades flagellin monomers. Here, we show that AprA is widespread among a wide variety of bacterial species. In addition, we investigated the role of AprA in virulence of the bacterial plant athogen P. syringae pv. tomato DC3000. The AprA-deficient DC3000 ΔaprA knockout mutant was significantl...

40 CFR 1400.5 - Internet access to certain off-site consequence analysis data elements.

Science.gov (United States)

2010-07-01

... consequence analysis data elements. 1400.5 Section 1400.5 Protection of Environment ENVIRONMENTAL PROTECTION... UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.5 Internet access to certain off...

Pain Perception: Computerized versus Traditional Local Anesthesia in Pediatric Patients.

Science.gov (United States)

Mittal, M; Kumar, A; Srivastava, D; Sharma, P; Sharma, S

2015-01-01

Local anesthetic injection is one of the most anxiety- provoking procedure for both children and adult patients in dentistry. A computerized system for slow delivery of local anesthetic has been developed as a possible solution to reduce the pain related to the local anesthetic injection. The present study was conducted to evaluate and compare pain perception rates in pediatric patients with computerized system and traditional methods, both objectively and subjectively. It was a randomized controlled study in one hundred children aged 8-12 years in healthy physical and mental state, assessed as being cooperative, requiring extraction of maxillary primary molars. Children were divided into two groups by random sampling - Group A received buccal and palatal infiltration injection using Wand, while Group B received buccal and palatal infiltration using traditional syringe. Visual Analog scale (VAS) was used for subjective evaluation of pain perception by patient. Sound, Eye, Motor (SEM) scale was used as an objective method where sound, eye and motor reactions of patient were observed and heart rate measurement using pulse oximeter was used as the physiological parameter for objective evaluation. Patients experienced significantly less pain of injection with the computerized method during palatal infiltration, while less pain was not statistically significant during buccal infiltration. Heart rate increased during both buccal and palatal infiltration in traditional and computerized local anesthesia, but difference between traditional and computerized method was not statistically significant. It was concluded that pain perception was significantly more during traditional palatal infiltration injection as compared to computerized palatal infiltration, while there was no difference in pain perception during buccal infiltration in both the groups.

Handbook of radiologic procedures

International Nuclear Information System (INIS)

Hedgcock, M.

1986-01-01

This book is organized around radiologic procedures with each discussed from the points of view of: indications, contraindications, materials, method of procedures and complications. Covered in this book are: emergency radiology chest radiology, bone radiology, gastrointestinal radiology, GU radiology, pediatric radiology, computerized tomography, neuroradiology, visceral and peripheral angiography, cardiovascular radiology, nuclear medicine, lymphangiography, and mammography

Functional similarities between the dictyostelium protein AprA and the human protein dipeptidyl-peptidase IV.

Science.gov (United States)

Herlihy, Sarah E; Tang, Yu; Phillips, Jonathan E; Gomer, Richard H

2017-03-01

Autocrine proliferation repressor protein A (AprA) is a protein secreted by Dictyostelium discoideum cells. Although there is very little sequence similarity between AprA and any human protein, AprA has a predicted structural similarity to the human protein dipeptidyl peptidase IV (DPPIV). AprA is a chemorepellent for Dictyostelium cells, and DPPIV is a chemorepellent for neutrophils. This led us to investigate if AprA and DPPIV have additional functional similarities. We find that like AprA, DPPIV is a chemorepellent for, and inhibits the proliferation of, D. discoideum cells, and that AprA binds some DPPIV binding partners such as fibronectin. Conversely, rAprA has DPPIV-like protease activity. These results indicate a functional similarity between two eukaryotic chemorepellent proteins with very little sequence similarity, and emphasize the usefulness of using a predicted protein structure to search a protein structure database, in addition to searching for proteins with similar sequences. © 2016 The Protein Society.

Functional similarities between the dictyostelium protein AprA and the human protein dipeptidyl‐peptidase IV

Science.gov (United States)

Herlihy, Sarah E.; Tang, Yu; Phillips, Jonathan E.

2017-01-01

Abstract Autocrine proliferation repressor protein A (AprA) is a protein secreted by Dictyostelium discoideum cells. Although there is very little sequence similarity between AprA and any human protein, AprA has a predicted structural similarity to the human protein dipeptidyl peptidase IV (DPPIV). AprA is a chemorepellent for Dictyostelium cells, and DPPIV is a chemorepellent for neutrophils. This led us to investigate if AprA and DPPIV have additional functional similarities. We find that like AprA, DPPIV is a chemorepellent for, and inhibits the proliferation of, D. discoideum cells, and that AprA binds some DPPIV binding partners such as fibronectin. Conversely, rAprA has DPPIV‐like protease activity. These results indicate a functional similarity between two eukaryotic chemorepellent proteins with very little sequence similarity, and emphasize the usefulness of using a predicted protein structure to search a protein structure database, in addition to searching for proteins with similar sequences. PMID:28028841

A Study on Conjugate Heat Transfer Analysis of Reactor Vessel including Irradiated Structural Heat Source

Energy Technology Data Exchange (ETDEWEB)

Yi, Kunwoo; Cho, Hyuksu; Im, Inyoung; Kim, Eunkee [KEPCO EnC, Daejeon (Korea, Republic of)

2015-10-15

Though Material reliability programs (MRPs) have a purpose to provide the evaluation or management methodologies for the operating RVI, the similar evaluation methodologies can be applied to the APR1400 fleet in the design stage for the evaluation of neutron irradiation effects. The purposes of this study are: to predict the thermal behavior whether or not irradiated structure heat source; to evaluate effective thermal conductivity (ETC) in relation to isotropic and anisotropic conductivity of porous media for APR1400 Reactor Vessel. The CFD simulations are performed so as to evaluate thermal behavior whether or not irradiated structure heat source and effective thermal conductivity for APR1400 Reactor Vessel. In respective of using irradiated structure heat source, the maximum temperature of fluid and core shroud for isotropic ETC are 325.8 .deg. C, 341.5 .deg. C. The total amount of irradiated structure heat source is about 5.41 MWth and not effect to fluid temperature.

The Numerical Sensitivity Study of Cold Leg Top Slot Break for ATLAS

Energy Technology Data Exchange (ETDEWEB)

Sung, Hae Jung; Lee, Sang Ik; Park, Ju hyun; Choi, Tong Soo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

In this study, cold leg top slot break calculations are performed by RELAP5/MOD3.3 Patch04 for the ATLAS test which is the scaled down experimental facility for the APR1400. The test condition for base case is selected as 0.1016 m (4 in.) based on the break size of the APR1400. In addition, sensitivity studies about break size, break distance from vessel, and pressurizer location are performed until the initiation of simultaneous injection of 2.5 hours. The loop seal reformation occurs early, and the duration of final loop seal reformation is longer as the break is close to vessel. Nonetheless, PCT increased by loop seal reformation is not identified since core uncovery does not occur. In this study, it is confirmed through RELAP5 simulation of the ATLAS test that cold leg top slot break for the APR1400 is not a safety issue in perspective of the loop seal reformation.

The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

International Nuclear Information System (INIS)

Jeong, Ji Hwan; Kweon, Young Chul

2002-01-01

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR 1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR 1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet

40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

Science.gov (United States)

2010-07-01

...-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL... PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.3 Public access to...

Reflecting on the ethical administration of computerized medical records

Science.gov (United States)

Collmann, Jeff R.

1995-05-01

This presentation examines the ethical issues raised by computerized image management and communication systems (IMAC), the ethical principals that should guide development of policies, procedures and practices for IMACS systems, and who should be involved in developing a hospital's approach to these issues. The ready access of computerized records creates special hazards of which hospitals must beware. Hospitals must maintain confidentiality of patient's records while making records available to authorized users as efficiently as possible. The general conditions of contemporary health care undermine protecting the confidentiality of patient record. Patients may not provide health care institutions with information about themselves under conditions of informed consent. The field of information science must design sophisticated systems of computer security that stratify access, create audit trails on data changes and system use, safeguard patient data from corruption, and protect the databases from outside invasion. Radiology professionals must both work with information science experts in their own hospitals to create institutional safeguards and include the adequacy of security measures as a criterion for evaluating PACS systems. New policies and procedures on maintaining computerized patient records must be developed that obligate all members of the health care staff, not just care givers. Patients must be informed about the existence of computerized medical records, the rules and practices that govern their dissemination and given the opportunity to give or withhold consent for their use. Departmental and hospital policies on confidentiality should be reviewed to determine if revisions are necessary to manage computer-based records. Well developed discussions of the ethical principles and administrative policies on confidentiality and informed consent and of the risks posed by computer-based patient records systems should be included in initial and continuing

Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Baek, W. P.; Song, C. H.; Kim, Y. S. and others

2005-02-15

The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

Discharge of 1400 AH lithium thionyl chloride cells into voltage reversal

Science.gov (United States)

Zolla, A. E.

1982-03-01

The development of a high energy density battery is discussed. Tests were conducted on a 1400 AH cell and they include: drop tests, crush tests, vibration and shock tests, hydrostatic pressure effects, charge current tests, and discharge tests. None of these tests led to a venting or explosion of the cell. Criteria are presented for extending RV safety of a 1400 AH cell.

Sonography and computerized tomography in a comparison in renal mass diseases

International Nuclear Information System (INIS)

Brommer, M.

1981-01-01

One hundred and thirteen patients with renal mass processes were examined sonographically and by computized tomography. The diagnosis was assured in 32 cases histologically, in 40 cases angiographically clinically and by the course and in 41 cases only clinically and by the course resp. by sonographic and computerized tomography control studies. The accuracy of a hit in sonography is smaller compared with computerized tomography with fast image formation. However, due to its relative safety, it is judged as the most useful technique for a first diagnossis of renal masses after ureographic diagnostic of excreta. A diagnostic procedure is proposed in which the number of necessary angiographies and diagnostic sampling surgery can be considerably reduced. (orig.) [de

A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

International Nuclear Information System (INIS)

Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

2012-01-01

A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

15 CFR 1400.3 - Request for determination.

Science.gov (United States)

2010-01-01

...) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.3... Business Development Agency, United States Department of Commerce, Washington, DC, 20230, containing the... detailed sociological, ethnic and/or racial description, as appropriate, of the group they represent which...

Interface of RETRAN/MASTER Code System for APR1400

International Nuclear Information System (INIS)

Ku, Keuk Jong; Kang, Sang Hee; Kim, Han Gon

2008-01-01

MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is a nuclear analysis and design code which can simulate the pressurized water reactor core or boiling water reactor core in 3-dimensional geometry. RETRAN is a best-estimate code for transient analysis of Non-LOCA. RETRAN code generates neutron number density in core using point kinetics model which includes feedback reactivities and converts the neutron number density into reactor power. It is conventional that RETRAN code for power generation is roughly to extrapolate feedback reactivities which are provided by MASTER code only one time before transient analysis. The purpose of this paper is to interface RETRAN code with MASTER code by real-time processing and to supply adequate feedback reactivities to RETRAN code. So, we develop interface code called MATRAN for real-time feedback reactivity processing. And for the application of MATRAN code, we compare the results of real-time MATRAN code with those of conventional RETRAN/MASTER code

Structural Analysis of Advanced Refueling Machine of APR1400

International Nuclear Information System (INIS)

Cho, J. R.; Kim, Y. H.; Park, B. T.; Park, J. B.; Jung, J. H.

2007-01-01

The Refueling Machine (RM) consists of two structural parts of bridge and trolley. The bridge structure is approximately 8.5 m long and 5 m wide and is primarily composed of two deep wide flange sections spanning the rector area at the operating level. The trolley is mounted on wheels that roll on the rails of the bridge. Vertical movements of trolley and bridge are restricted by guide rollers. In this paper, dynamic and structural analyses based on the earthquake spectrum are carried out to verify the structural integrity of advanced refueling machine. It is done by 3-dimensional finite element analysis using ANSYS software

DECA -- The design change assistant: An application of expert systems concepts to procedure automation

International Nuclear Information System (INIS)

Brtis, J.S.

1993-01-01

Proceduralized engineering reviews required for nuclear power stations result in significant O ampersand M cost's. A method of streamlining procedural reviews (while improving their quality) is to computerize the procedures and associated checklists. DECA, the Design Change Assistant, is an expert system based program that is specifically designed for computerizing procedures. This program improves the effectiveness of engineers in performing engineering design reviews. DECA is a development shell, which allows the computerization of most types of engineering reviews. In addition to computerizing the decision making process, DECA makes it possible to call on electronic information, such as databases or text files, and makes them available to the user to assist in a review. DECA clearly demonstrates the benefits that can accrue by applying expert system technologies to engineering reviews. It results in the following: a well defined process, reduced user training in the use of procedures, improved quality of reviews, a streamlining of the review process through elimination of questions that are not applicable, and high quality documentation of reviews. DECA has been successfully applied to Commonwealth Edison's engineering design review checklists, and is currently in use by Commonwealth Edison

Could conscious sedation with midazolam for dental procedures be ...

African Journals Online (AJOL)

2012-04-25

Apr 25, 2012 ... they used CS. Minor oral surgical procedures and tooth extraction processes requiring no saline irrigation, however, ... The pre-diagnosis and related treatment plans .... accompanying tooth decay; orthodontic tooth extractions.

A New Human-Machine Interfaces of Computer-based Procedure System to Reduce the Team Errors in Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Sa Kil; Sim, Joo Hyun; Lee, Hyun Chul

2016-01-01

In this study, we identify the emerging types of team errors, especially, in digitalized control room of nuclear power plants such as the APR-1400 main control room of Korea. Most works in nuclear industry are to be performed by a team of more than two persons. Even though the individual errors can be detected and recovered by the qualified others and/or the well trained team, it is rather seldom that the errors by team could be easily detected and properly recovered by the team itself. Note that the team is defined as two or more people who are appropriately interacting with each other, and the team is a dependent aggregate, which accomplishes a valuable goal. Organizational errors sometimes increase the likelihood of operator errors through the active failure pathway and, at the same time, enhance the possibility of adverse outcomes through defensive weaknesses. We incorporate the crew resource management as a representative approach to deal with the team factors of the human errors. We suggest a set of crew resource management training procedures under the unsafe environments where human errors can have devastating effects. We are on the way to develop alternative interfaces against team error in a condition of using a computer-based procedure system in a digitalized main control room. The computer-based procedure system is a representative feature of digitalized control room. In this study, we will propose new interfaces of computer-based procedure system to reduce feasible team errors. We are on the way of effectiveness test to validate whether the new interface can reduce team errors during operating with a computer-based procedure system in a digitalized control room

A New Human-Machine Interfaces of Computer-based Procedure System to Reduce the Team Errors in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Sa Kil; Sim, Joo Hyun; Lee, Hyun Chul [Korea Atomic Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this study, we identify the emerging types of team errors, especially, in digitalized control room of nuclear power plants such as the APR-1400 main control room of Korea. Most works in nuclear industry are to be performed by a team of more than two persons. Even though the individual errors can be detected and recovered by the qualified others and/or the well trained team, it is rather seldom that the errors by team could be easily detected and properly recovered by the team itself. Note that the team is defined as two or more people who are appropriately interacting with each other, and the team is a dependent aggregate, which accomplishes a valuable goal. Organizational errors sometimes increase the likelihood of operator errors through the active failure pathway and, at the same time, enhance the possibility of adverse outcomes through defensive weaknesses. We incorporate the crew resource management as a representative approach to deal with the team factors of the human errors. We suggest a set of crew resource management training procedures under the unsafe environments where human errors can have devastating effects. We are on the way to develop alternative interfaces against team error in a condition of using a computer-based procedure system in a digitalized main control room. The computer-based procedure system is a representative feature of digitalized control room. In this study, we will propose new interfaces of computer-based procedure system to reduce feasible team errors. We are on the way of effectiveness test to validate whether the new interface can reduce team errors during operating with a computer-based procedure system in a digitalized control room.

Assumptions and Policy Decisions for Vital Area Identification Analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Myungsu; Bae, Yeon-Kyoung; Lee, Youngseung [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

U.S. Nuclear Regulatory Commission and IAEA guidance indicate that certain assumptions and policy questions should be addressed to a Vital Area Identification (VAI) process. Korea Hydro and Nuclear Power conducted a VAI based on current Design Basis Threat and engineering judgement to identify APR1400 vital areas. Some of the assumptions were inherited from Probabilistic Safety Assessment (PSA) as a sabotage logic model was based on PSA logic tree and equipment location data. This paper illustrates some important assumptions and policy decisions for APR1400 VAI analysis. Assumptions and policy decisions could be overlooked at the beginning stage of VAI, however they should be carefully reviewed and discussed among engineers, plant operators, and regulators. Through APR1400 VAI process, some of the policy concerns and assumptions for analysis were applied based on document research and expert panel discussions. It was also found that there are more assumptions to define for further studies for other types of nuclear power plants. One of the assumptions is mission time, which was inherited from PSA.

Mid-Infrared Observations of the White Dwarf Brown Dwarf Binary GD 1400

OpenAIRE

Farihi, J.; Zuckerman, B.; Becklin, E. E.

2005-01-01

Fluxes are measured for the DA white dwarf plus brown dwarf pair GD 1400 with the Infrared Array Camera on the {\\em Spitzer Space Telescope}. GD 1400 displays an infrared excess over the entire $3-8\\mu$m region consistent with the presence of a mid- to late-type L dwarf companion. A discussion is given regarding current knowledge of this unique system.

48 CFR 1552.235-76 - Treatment of Confidential Business Information (APR 1996).

Science.gov (United States)

2010-10-01

... Business Information (APR 1996). 1552.235-76 Section 1552.235-76 Federal Acquisition Regulations System... Provisions and Clauses 1552.235-76 Treatment of Confidential Business Information (APR 1996). As prescribed in 1535.007-70(c), insert the following clause: Treatment of Confidential Business Information (TSCA...

Computerizing primary schools in rural kenya

DEFF Research Database (Denmark)

Ogembo, J.G.; Ngugi, B.; Pelowski, Matthew John

2012-01-01

questions surrounding this endeavour. Specifically: 1.) what problems do rural schools actually want to solve with computerization; 2.) is computerization the most important priority for rural schools; 3.) are schools ready, in terms of infrastructure, for a computer in the classroom; or 4.) might...... and protective roofing -posing severe challenges to the outstanding conception of computerization. We consider these results and make recommendations for better adapting programs for computer introduction, and also suggest the use of new innovative devices, such as cell phones, which might already have overcome......This paper investigates the outstanding challenges facing primary schools' computerization in rural Kenya. Computerization of schools is often envisaged as a 'magic', or at least a particularly efficient, solution to many of the problems that developing countries face in improving primary school...

Computerized transverse tomography of vascular lesions of the brain. II. Aneurysms

International Nuclear Information System (INIS)

Pressman, B.D.; Gilbert, G.E.; Davis, D.O.

1975-01-01

Computerized transverse tomography (CTT) is an excellent screening examination for intracerebral disease. Unfortunately, the detection rate of small aneurysms by this procedure is not high. Nevertheless, CTT may be useful in aneurysm cases since it may indicate the need for further evaluation in patients with associated hematomas, encephalomalacia, or ventricular dilatation, even if the aneurysm itself is not delineated. (U.S.)

Implementasi Automatic Packet Reporting System (APRS) Untuk Paket Data Pemantauan dan PengukuranUntuk Paket Data Pemantauan dan Pengukuran

OpenAIRE

Arief Goeritno

2016-01-01

Telah dilakukan implementasi Automatic Packet Reporting System (APRS) untuk paket data pemantauan dan pengukuran melalui tujuan penelitian, berupa: a) penyetelan program aplikasi pada jaringan APRS dan b) pengukuran terhadap penerimaan data berdasarkan kinerja sensor-sensor. Penyetelan program aplikasi APRS merupakan konfigurasi perangkat lunak untuk APRS yang akan digunakan pada stasiun penerimaan data APRS dengan program aplikasi yang biasa digunakan, yaitu hyperterminal dan UI-View 32. Pen...

Computerized real-time materials accountability system for safeguards material control

International Nuclear Information System (INIS)

Spencer, W.F.; Affel, R.G.; Austin, H.C.; Nichols, J.P.; Stoutt, B.H.; Wachter, J.W.

1975-01-01

A real-time, computer-based system is described which provides safeguards material control at the Oak Ridge National Laboratory. Originally installed in 1972 to provide computerized real-time fissile materials accountability for criticality control purposes, the system has been expanded to provide accountability of all source and nuclear materials (SNM) and to utilize the on-line inventory files in support of the Laboratory physical protection and surveillance procedures. (auth)

Computer assisted procedure maintenance

International Nuclear Information System (INIS)

Bisio, R.; Hulsund, J. E.; Nilsen, S.

2004-04-01

The maintenance of operating procedures in a NPP is a tedious and complicated task. Through the whole life cycle of the procedures they will be dynamic, 'living' documents. Several aspects of the procedure must be considered in a revision process. Pertinent details and attributes of the procedure must be checked. An organizational structure must be created and responsibilities allotted for drafting, revising, reviewing and publishing procedures. Available powerful computer technology provides solutions within document management and computerisation of procedures. These solutions can also support the maintenance of procedures. Not all parts of the procedure life cycle are equally amenable to computerized support. This report looks at the procedure life cycle in todays NPPs and discusses the possibilities associated with introduction of computer technology to assist the maintenance of procedures. (Author)

Design and Analysis for an Operator's Action Sharing System

Energy Technology Data Exchange (ETDEWEB)

Jung, Yeonsub; Seong, Nokyu [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2013-10-15

In the paper, contrary to HRP approach, no more FPD is introduced because there is no room to install. The same workstation and the same LDP should be utilized for ream transparency. A action sharing system will be introduced at ShinKori 3,4 MCR, and further applied to other APR1400 plant if necessary. The project started and applied by the end of 2014. Despite benefit of action sharing system, there are lots of challenges to overcome such as traffic load, and interfaces. The challenges have been analyzed thoroughly. Traffic load can be reduced through vector graphics, video driver, and capture and compressing techniques. Furthermore interfaces for action sharing system are developed and evaluated to reduce secondary workload. Advanced digital control rooms have lots of advantages compared to analog control room. They can integrate all process variables into more comprehensible forms. Advanced alarm processor can suppress trivial alarms, and P and ID based mimic isplays can be integrated with context sensitive menu for referencing. Moreover computer based procedures have been introduced at more advanced MCR. Because all these display appears at flat panel display (FPD), they can be easily modified if necessary. These days newly introduced MCRs are advanced types, and analog control rooms are no more built. In spite of this trend, advanced control rooms have shortage in view of team transparency. For example, shift supervisor cannot tell which devices reactor operator is manipulating. APR1400 MCR has large display panel to share the same situation awareness among crew member. Because LDP has fixed display comparing switchable display in FPD, situation awareness can be enhanced. However, even LDP cannot show the active device that crew member are manipulating due to either limited number of devices in LDP or no demarcation for the active device. During construction of ShinKori 3/4, the demarcation box for the active device has been introduced and called an Active

Design and Analysis for an Operator's Action Sharing System

International Nuclear Information System (INIS)

Jung, Yeonsub; Seong, Nokyu

2013-01-01

In the paper, contrary to HRP approach, no more FPD is introduced because there is no room to install. The same workstation and the same LDP should be utilized for ream transparency. A action sharing system will be introduced at ShinKori 3,4 MCR, and further applied to other APR1400 plant if necessary. The project started and applied by the end of 2014. Despite benefit of action sharing system, there are lots of challenges to overcome such as traffic load, and interfaces. The challenges have been analyzed thoroughly. Traffic load can be reduced through vector graphics, video driver, and capture and compressing techniques. Furthermore interfaces for action sharing system are developed and evaluated to reduce secondary workload. Advanced digital control rooms have lots of advantages compared to analog control room. They can integrate all process variables into more comprehensible forms. Advanced alarm processor can suppress trivial alarms, and P and ID based mimic isplays can be integrated with context sensitive menu for referencing. Moreover computer based procedures have been introduced at more advanced MCR. Because all these display appears at flat panel display (FPD), they can be easily modified if necessary. These days newly introduced MCRs are advanced types, and analog control rooms are no more built. In spite of this trend, advanced control rooms have shortage in view of team transparency. For example, shift supervisor cannot tell which devices reactor operator is manipulating. APR1400 MCR has large display panel to share the same situation awareness among crew member. Because LDP has fixed display comparing switchable display in FPD, situation awareness can be enhanced. However, even LDP cannot show the active device that crew member are manipulating due to either limited number of devices in LDP or no demarcation for the active device. During construction of ShinKori 3/4, the demarcation box for the active device has been introduced and called an Active

40 CFR 1400.6 - Enhanced local access.

Science.gov (United States)

2010-07-01

... INFORMATION Public Access § 1400.6 Enhanced local access. (a) OCA data elements. Consistent with 42 U.S.C..., may convey to the public OCA data elements orally or in writing, as long as the data elements are not... Administrator from those sections. Disseminating OCA data elements to the public in a manner consistent with...

Patient surface doses in computerized tomography examinations

International Nuclear Information System (INIS)

Vekic, B; Kovacevic, S.; Ranogajec-Komor, M.; Duvnjak, N.; Marusic, P.; Anic, P.; Dolencic, P.

1996-01-01

The diagnostic value of computerized tomography has increased due to very rapid technical advances in both equipment and techniques. When the CT scanners were introduced, a significant problem for the specification of the radiation dose imparted to the patient undergoing CT examination has been created. In CT, the conditions of exposure are quite different from those in conventional X-ray imaging. CT procedure involves the continuous tomography of thin layers. Some of these layers touch each other while others overlap. The radiation doses received by patients can vary considerably. In addition to the radiation from the collimated primary beam, patients are exposed to significant scattered doses in unpredictable amounts. Every effort should be made to keep these doses to a reasonable minimum, without sacrificing the image quality. The aims of this work were to determine the surface doses delivered to various organs of patients during various computerized tomography examinations (head, thorax, kidney, abdomen and pelvis). Particular attention was directed to the precise determination of doses received by the eyes (during CT of head) and gonads (during CT of pelvis and lower abdomen) since these organs can be near or even in the primary X-ray beam

48 CFR 1552.235-75 - Access to Toxic Substances Control Act Confidential Business Information (APR 1996).

Science.gov (United States)

2010-10-01

... Control Act Confidential Business Information (APR 1996). 1552.235-75 Section 1552.235-75 Federal... Confidential Business Information (APR 1996). As prescribed in 1535.007(b), insert the following provision: Access to Toxic Substances Control Act Confidential Business Information (APR 1996) In order to perform...

Variability of nursing care by APR-DRG and by severity of illness in a sample of nine Belgian hospitals.

Science.gov (United States)

Pirson, Magali; Delo, Caroline; Di Pierdomenico, Lionel; Laport, Nancy; Biloque, Veronique; Leclercq, Pol

2013-10-10

As soon as Diagnosis related Groups (DRG) were introduced in many hospital financing systems, most nursing research revealed that DRG were not very homogeneous with regard to nursing care. However, few studies are based on All Patient refined Diagnosis related Groups (APR-DRGs) and few of them use recent data. Objectives of this study are: (1) to evaluate if nursing activity is homogeneous by APR-DRG and by severity of illness (SOI) (2) to evaluate the outlier's rate associated with the nursing activity and (3) to compare nursing cost homogeneity per DRG and SOI. Study done in 9 Belgian hospitals on a selection of APR-DRG with more than 30 patients (7 638 inpatient stays). The evaluation of the homogeneity is based on coefficients of variation (CV). The 75th percentile + 1.5 × inter-quartile range was used to select high outliers. 25th percentile -1.5 × inter-quartile range was used to select low outliers. Nursing costs per ward were distributed on inpatient stays of each ward following two techniques (the LOS vs. the number of nursing care minutes per stay). The homogeneity of LOS by DRG and by SOI is relatively good (CV: 0.56). The homogeneity of the nursing activity by DRG is less good (CVs between 0.36 and 1.54) and is influenced by nursing activity outliers (high outliers' rate: 5.19%, low outliers' rate: 0.14%). The outlier's rate varies according to the studied variable. The high outliers' rate is higher for nursing activity than for LOS. The homogeneity of nursing costs is higher when costs are based on the LOS of patients than when based on minutes of nursing care (CVs between 0.26 and 1.46 for nursing costs based on LOS and between 0.49 and 2.04 for nursing costs based on minutes of nursing care). It is essential that the calculation of nursing cost by stay and by DRG for hospital financing purposes was based on nursing activity data, that more reflect resources used in wards, and not on LOS data. The only way to obtain this information is

Computerized adaptive testing in computer assisted learning?

NARCIS (Netherlands)

Veldkamp, Bernard P.; Matteucci, Mariagiulia; Eggen, Theodorus Johannes Hendrikus Maria; De Wannemacker, Stefan; Clarebout, Geraldine; De Causmaecker, Patrick

2011-01-01

A major goal in computerized learning systems is to optimize learning, while in computerized adaptive tests (CAT) efficient measurement of the proficiency of students is the main focus. There seems to be a common interest to integrate computerized adaptive item selection in learning systems and

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

International Nuclear Information System (INIS)

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

2006-01-01

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

The computerized OMAHA system in microsoft office excel.

Science.gov (United States)

Lai, Xiaobin; Wong, Frances K Y; Zhang, Peiqiang; Leung, Carenx W Y; Lee, Lai H; Wong, Jessica S Y; Lo, Yim F; Ching, Shirley S Y

2014-01-01

The OMAHA System was adopted as the documentation system in an interventional study. To systematically record client care and facilitate data analysis, two Office Excel files were developed. The first Excel file (File A) was designed to record problems, care procedure, and outcomes for individual clients according to the OMAHA System. It was used by the intervention nurses in the study. The second Excel file (File B) was the summary of all clients that had been automatically extracted from File A. Data in File B can be analyzed directly in Excel or imported in PASW for further analysis. Both files have four parts to record basic information and the three parts of the OMAHA System. The computerized OMAHA System simplified the documentation procedure and facilitated the management and analysis of data.

Examinations and patient management in radiologic diagnostics by means of a computerized data-processing communications system

International Nuclear Information System (INIS)

Klotz, E.; Remplik, V.; Opfer, M.; Wilde, E.

1986-01-01

Concept, realization and a two-year experience in routine use of a computerized data-processing communications system supporting the central radiology department in a 1.625-bed, municipal hospital are discussed. Via the hospital ward-linked terminal network, routine X-ray examination is so ordered, that individual medical case information is provided, which certifies qualified clinical and diagnostic procedure along with good patient preparation and transportation. Complete information and documentation are also ensured. It was shown, that good motivation among involved radiologists and their close working-association with other departments are mandatory from the start to maintain exact time schedules and satisfactory completion of ordered X-ray procedures. Subsequently, the assistance through computerized data-processing becomes an accepted help - so far as it is appropriately used - as time goes by in its routine use. (orig.) [de

A Study on the Risk Reduction Effect by MLCS (Mid-loop Level Control System) of EUAPR using the Low-Power and Shutdown PSA Result

International Nuclear Information System (INIS)

Lee, Keunsung; Choi, Sunmi; Kim, Eden

2016-01-01

The EU-APR design has been developed in order to expand and diversify the global nuclear power market of APR1400. For the improvement of shutdown risk for the EUAPR, the mid-loop level control system (MLCS) is considered during mid-loop operation for the EU-APR, which is not incorporated into SKN 3 and 4 (APR1400 Type) in Korea. Commonly, the risk associated with the NPP can be identified through the PSA. Thus, this paper discusses the low power and shutdown (LPSD) risk reduction effect by MLCS using the Low-Power and Shutdown PSA Result. LPSD level 1 PSA models for EU-APR have been developed. The risk reduction effect by MLCS is discussed. Because the loss of shutdown cooling function during mid-loop is one of the most vulnerable events, the MLCS have a significant influence on CDF in LPSD PSA. The shutdown risk of domestic power plants would likely be reduced if the MLCS is adopted in all operating NPPs in Korea during the mid-loop operation. It is expected that this work will contribute to reduce shutdown risk of domestic power plants

40 CFR 1400.4 - Vulnerable zone indicator system.

Science.gov (United States)

2010-07-01

....4 Section 1400.4 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7... further information. (b) Methods of access. The indicator shall be available on the Internet or by request...

Feasibility study for a computerized emergency preparedness simulation facility

International Nuclear Information System (INIS)

Gerhardstein, L.H.; Schroeder, J.O.; Sandusky, W.F.

1979-11-01

This report details the feasibility of a computerized Emergency Preparedness Simulation Facility (EPSF) for use by the Nuclear Regulatory Commission (NRC). The proposed facility would be designed to provide the NRC and other federal, state, and local government agencies with a capability to formulate, test, and evaluate the Emergency Preparedness Plans (EPP) which local and state agencies have/will establish for use during nuclear emergencies. In cases of any state emergency (including a nuclear emergency), high level state government officials will direct emergency procedures and insure that state and local emergency teams carry out tasks which have been established in their EPP. When an emergency exists, rapid mobilization of emergency teams, efficient communication, and effective coordination of individual team efforts is essential to safety, preservation of property, and overall public welfare. Current EPP evaluation procedures are qualitative in nature and while they do compare emergency drill performance with the EPP, the nature of the drills often does not provide enough realism to actual emergency conditions. Automated simulation of real emergency conditions using modern computer equipment and programming techniques will provide the NRC emergency evaluation teams a simulated environment which closely approximates conditions which would actually exist during a real emergency. In addition, the computer can be used to collect and log performance and event data which will aid the evaluation team in making assessments of the state or local area's EPP and their Emergency Preparedness Teams performance during emergency drills. Overall, a computerized EPSF can improve drill testing and evaluation efficiency, provide approximate emergency condition realism, and improve public awareness of local emergency procedures

Design development of steel plate concrete modularization for the advanced PWR in Korea

International Nuclear Information System (INIS)

Mun, Taeyoup; Kim, Keunkyeong; Sun, Wonsang; Kim, Taeyoung; Hwang, Geunha

2008-01-01

APR1400 TM - an advanced PWR - has been developed in Korea since 1992. Four APR1400 units - Shin Kori no.3,4 and Shin Uljin no.1,2 - are going to be built in a next decade. As for economical efficiency, construction cost per power generation Unit(W) is improved more than 10% compared to the former 1,000 MWe PWRs. Moreover, life-cycle maintenance cost is reduced to the world's most level. For construction period from first concrete pouring to commercial operation, 54 months for APR1400 and 36 months for n-th unit have been projected. Reduction of the construction term of the Nuclear Power Plant has been emphasized increasingly for the NPP construction Project because it would reduce the interest cost and uncertainty of the project. The reduction can also advance the return of investment. Some of the PPM(Prefabrication, Preassembly, and Modularization) techniques have been studied for the shortening the construction period of nuclear power plant. Especially for the internal structure of reactor containment building (RCB) in PWR whose term of construction is critical to the whole project, Steel Plate Concrete(SC) structure has been studied as one of alternative structural systems to the conventional Reinforced Concrete(RC) structure in APR1400. SC structure is considered appropriate for the modularization of the structure with its self-supporting. In addition, formwork can be dramatically eliminated when SC structural modules are used. The MKE (Ministry of Knowledge Economy) and KHNP (Korea Hydro and Nuclear Power Co., Ltd.) initiated the research and development of SC Structure in 2005. This paper presents design examples along with Codes and Standards of SC structure in nuclear power plant. (author)

Performance Test of BF3 Neutron Detection System

Energy Technology Data Exchange (ETDEWEB)

Choi, Yu Sun; Shin, Ho Cheol [KHNP-CRI, Daejeon (Korea, Republic of); Cho, Jin Bok; Oh, Sae Hyun; Ryou, Seok Jean [USERS, Daejeon (Korea, Republic of)

2015-10-15

The neutron detecting system of First-of-a-kind plant such an APR1400 at Shin Kori should have been verified in the condition of low operating temperature and pressure of the primary coolant system before receiving the operation license. Auxiliary Ex-core Neutron Flux Monitoring System (AENFMS) is supposed to be installed using BF3 neutron detector in Shin Kori plant. The performance test of AENFMS was conducted to measure neutron sensitivity, moderation ratio and count rate in the same condition with Ex-core Neutron Flux Monitoring System (ENFMS) of APR1400 to verify its detection characteristics in compliance with the functional requirement. Performance test has been conducted for AENFMS of APR1400 to verify BF3 neutron sensitivity, moderation ration of PE, expecting neutron signal count rate from AENFMS, possible extending cable length from detector to pre-amplifier. As a result of measurement, the neutron sensitivity of 34.246±0.168(95%CI)cps/nv, moderation ratio of 11.343±0.039(95%CI) and AENFMS expecting count rate related to ENFMS of 17.8 times are acceptable in compliance with functional requirement, respectively.

A Proposal to upgrade KREM with respect to Scale and Code Deficiency

International Nuclear Information System (INIS)

Lee, Sang Yong

2015-01-01

In this paper, it is shown that the EDC process can be formulated with respect to the scaled system uncertainty quantification. The second suggestion is concerning the bias due to the scale deficiency related with the multi-dimensional effect. A proposal to extend KREM to have multi-dimensional capability is made so that the related bias can be removed. Korea Realistic Evaluation Model (KREM) basically follows Code Scale Applicability and Uncertainty (CSAU) methodology (Fig.1 and 2). Since KREM was approved by the Korean authority in 2002, it has been applied to support several important domestic and foreign projects such as APR-1400(Advanced Power Reactor) design, APR-1400 export to United Arab Emirates (UAE) and U.S. NRC design certification of APR-1400. Encouraged by these activities, I would like to propose some suggestions to modify and improve KREM. The first suggestion is related with the Experimental Data Covering (EDC) process (Fig.2). The terminology SSUQ is introduced to replace the term EDC in KREM to naturally include the operational parameters. COBRA-TF is strongly recommended to handle the multi-dimensional phenomena in the improved KREM

Computerized tomography-guided neurolytic splanchnic nerve block

International Nuclear Information System (INIS)

Henriquet, Franco; De Martini, Giuseppe; Roy, Maria Teresa; Pretrolesi, Fabio; Martinoli, Carlo; Cariati, Maurizio; Fiorentini, Franco.

1997-01-01

Computerized tomography-guided neurolytic splanchnic nerve block is a technique for relieving abdominal cancer pain; the goal is the alcoholic neurolytic interruption of the sensitive structures in retroperitoneal space. Computerized tomography yields accurate anatomical detailing and the course for needle placement and alcohol spread. January, 1993, to July, 1996, twenty-one bilateral splanchnic nerve blocks were performed through the posterior access. Forty-eight hours after alcoholism. 14 patients (66%) had complete pain regression; 52% of the patients needed no analgesics for 6 to 54 days and only 9 patients (42%) needed another low opioid therapy. Complications included hypotension and diarrhea in all cases. One had a cardiac arrest and diet 8 days after the procedure. There were no other complications. The whole procedure usually lasted 60 min (range: 45 to 90 min). Splanchnic nerve neurolysis is a useful treatment in the patients with severe chronic abdominal pain. It is used as a second line treatment when large lesions change celia anatomy and complicate the percutaneous block of the celiac plexus. Endosulfan, Malathion and Methyl parathion, on the metabolic rate of the estuarine clam, Villorita cyprinoides var. cochinensis, have been investigated. The animals exposed to the lower sublethal concentrations of Endosulfan, Malthion and Methyl parathion consumed oxygen at the rate of 1.60, 1.98 and 2.09 ml. 0 2 g - 1 h -1 respectively, while at the higher concentrations of the pesticides, consumption of oxygen by the animal dropped to nearly half the control value. When compared to Malathion and Methyl parathion. Endosulfan induced animals recorded a greater reduction in her percentage deviation (from control) of oxygen consumption, possibly due to hypoxia induced by the pollutants

Prediction Capability of SPACE Code about the Loop Seal Clearing on ATLAS SBLOCA

Energy Technology Data Exchange (ETDEWEB)

Bae, Sung Won; Lee, Jong Hyuk; Chung, Bub Dong; Kim, Kyung Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The most possible break size for loop seal reforming has been decided as 4 inch by the pre-calculation conducted by the RELAP5 and MARS. Many organizations have participated with various system analysis codes: for examples, RELAP5, MARS, TRACE. KAERI also anticipated with SPACE code. SPACE code has been developed for the use of design and safety analysis of nuclear thermal hydraulics system. KHNP and other organizations have collaborated during last 10 years. And it is currently under the certification procedures. SPACE has the capability to analyze the droplet field with full governing equation set: continuity, momentum, and energy. The SPACE code has been participated in PKL- 3 benchmark program for the international activity. The DSP-04 benchmark problem is also the application of SPACE as the domestic activities. The cold leg top slot break accident of APR1400 reactor has been modeled and surveyed by SPACE code. Benchmark experiment as a program of DSP-04 has been performed with ATLAS facility. The break size has been selected as 4 inch in APR1400 and the corresponding scale down break size has been modeled in SPACE code. The loop seal reforming has been occurred at all 4 loops. But the PCT shows no significant behaviors.

Computerized Adaptive Personality Testing: A Review and Illustration With the MMPI-2 Computerized Adaptive Version.

Science.gov (United States)

Forbey, Johnathan D.; Ben-Porath, Yossef S.

2007-01-01

Computerized adaptive testing in personality assessment can improve efficiency by significantly reducing the number of items administered to answer an assessment question. Two approaches have been explored for adaptive testing in computerized personality assessment: item response theory and the countdown method. In this article, the authors…

Strategy generator in computerized accident management support system

International Nuclear Information System (INIS)

Sirola, M.

1994-02-01

An increased interest for research in the field of accident management of nuclear power plants can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accidents in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The idea of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information from the plant will help the strategy planning. (orig.). (40 refs., 20 figs.)

Business environment of nuclear power industry in Korea

International Nuclear Information System (INIS)

Lee, Yoon Young

2003-01-01

In Korea, there are total of 18 Nuclear Power Plants in operation as of the end of 2002 and 6 more plants are under construction. The first project for the Advanced Power Reactor (APR) 1400 nuclear power plant is being launched to provide reliable electricity economical competitiveness in Korea. The competitive business environment both globally and in Korea, where the power industry is undergoing significant restructuring, is requiring the Korean nuclear industry to continually improve the economic associated with nuclear power. Introduction of the APR 1400 design and continued improvement of local capabilities are two of the ways that the industry is responding to the challenge. (author)

Insecticidal activity of the metalloprotease AprA occurs through suppression of host cellular and humoral immunity.

Science.gov (United States)

Lee, Seung Ah; Jang, Seong Han; Kim, Byung Hyun; Shibata, Toshio; Yoo, Jinwook; Jung, Yunjin; Kawabata, Shun-Ichiro; Lee, Bok Luel

2018-04-01

The biochemical characterization of virulence factors from entomopathogenic bacteria is important to understand entomopathogen-insect molecular interactions. Pseudomonas entomophila is a typical entomopathogenic bacterium that harbors virulence factors against several insects. However, the molecular actions of these factors against host innate immune responses are not clearly elucidated. In this study, we observed that bean bugs (Riptortus pedestris) that were injected with P. entomophila were highly susceptible to this bacterium. To determine how P. entomophila counteracts the host innate immunity to survive within the insect, we purified a highly enriched protein with potential host insect-killing activity from the culture supernatant of P. entomophila. Then, a 45-kDa protein was purified to homogeneity and identified as AprA which is an alkaline zinc metalloprotease of the genus Pseudomonas by liquid chromatography mass spectrometry (LC-MS). Purified AprA showed a pronounced killing effect against host insects and suppressed both host cellular and humoral innate immunity. Furthermore, to show that AprA is an important insecticidal protein of P. entomophila, we used an aprA-deficient P. entomophila mutant strain (ΔaprA). When ΔaprA mutant cells were injected to host insects, this mutant exhibited extremely attenuated virulence. In addition, the cytotoxicity against host hemocytes and the antimicrobial peptide-degrading ability of the ΔaprA mutant were greatly decreased. These findings suggest that AprA functions as an important insecticidal protein of P. entomophila via suppression of host cellular and humoral innate immune responses. Copyright © 2017 Elsevier Ltd. All rights reserved.

Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

International Nuclear Information System (INIS)

Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

2010-01-01

This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

Integration of computerized operation support systems on a nuclear power plant environment

Energy Technology Data Exchange (ETDEWEB)

Jaime, Guilherme D.G.; Almeida, Jose C.S.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.br, E-mail: jcsa@ien.gov.br, E-mail: mvitor@ien.gov.br [Instituto Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Servico de Sistemas Complexos. Divisao de Engenharia Nuclear

2015-07-01

Automation of certain tasks in a Nuclear Power Plant (NPP) control room is expected to result in reduced operators' mental workload, which may induce other benefits such as enhanced situation awareness and improved system performance. The final goal should be higher level of operational safety. Thus, recent works are increasingly assessing automation. The LABIHS compact NPP simulator, though, still operates under strictly manual printed hard-copy procedures, despite of the fact that the simulator incorporates several advancements in design of digitalized Human-Interfaces (HSIs). This work presents the development, implementation and integration of selected components to achieve increased level of computerized/automated operation of the LABIHS compact NPP simulator. Specifically, we discuss three components: (I) Automatic Plant Mode Detection, (II) Automatic Alarm Filtering, and (III) Computerized Procedures. Each one of these components has to be carefully designed/integrated so that one can avoid the undesired effects of some known implementations of automated systems on NPP, such as the reduction in the operator's system awareness, an increase in monitoring workload, and the degradation in manual skills, which could lead to automation-induced system failures. (author)

Integration of computerized operation support systems on a nuclear power plant environment

International Nuclear Information System (INIS)

Jaime, Guilherme D.G.; Almeida, Jose C.S.; Oliveira, Mauro V.

2015-01-01

Automation of certain tasks in a Nuclear Power Plant (NPP) control room is expected to result in reduced operators' mental workload, which may induce other benefits such as enhanced situation awareness and improved system performance. The final goal should be higher level of operational safety. Thus, recent works are increasingly assessing automation. The LABIHS compact NPP simulator, though, still operates under strictly manual printed hard-copy procedures, despite of the fact that the simulator incorporates several advancements in design of digitalized Human-Interfaces (HSIs). This work presents the development, implementation and integration of selected components to achieve increased level of computerized/automated operation of the LABIHS compact NPP simulator. Specifically, we discuss three components: (I) Automatic Plant Mode Detection, (II) Automatic Alarm Filtering, and (III) Computerized Procedures. Each one of these components has to be carefully designed/integrated so that one can avoid the undesired effects of some known implementations of automated systems on NPP, such as the reduction in the operator's system awareness, an increase in monitoring workload, and the degradation in manual skills, which could lead to automation-induced system failures. (author)

Review of APR+ Level 2 PSA

International Nuclear Information System (INIS)

Lehner, J.R.; Mubayi, V.; Pratt, W.T.

2012-01-01

Brookhaven National Laboratory (BNL) assisted the Korea Institute of Nuclear Safety (KINS) in reviewing the Level 2 Probabilistic Safety Assessment (PSA) of the APR+ Advanced Pressurized Water Reactor (PWR) prepared by the Korea Hydro and Nuclear Power Co., Ltd (KHNP) and KEPCO Engineering and Construction Co., Inc. (KEPCO-E and C). The work described in this report involves a review of the APR+ Level 2 PSA submittal (Ref. 1). The PSA and, therefore, the review is limited to consideration of accidents initiated by internal events. As part of the review process, the review team also developed three sets of Requests for Additional Information (RAIs). These RAIs were provided to KHNP and KEPCO-E and C for their evaluation and response. This final detailed report documents the review findings for each technical element of the PSA and includes consideration of all of the RAIs made by the reviewers as well as the associated responses. This final report was preceded by an interim report (Ref. 2) that focused on identifying important issues regarding the PSA. In addition, a final meeting on the project was held at BNL on November 21-22, 2011, where BNL and KINS reviewers discussed their preliminary review findings with KHNP and KEPCO-E and C staffs. Additional information obtained during this final meeting was also used to inform the review findings of this final report. The review focused not only on the robustness of the APR+ design to withstand severe accidents, but also on the capability and acceptability of the Level 2 PSA in terms of level of detail and completeness. The Korean nuclear regulatory authorities will decide whether the PSA is acceptable and the BNL review team is providing its comments for KINS consideration. Section 2.0 provides the basis for the BNL review. Section 3.0 presents the review of each technical element of the PSA. Conclusions and a summary are presented in Section 4.0. Section 5.0 contains the references.

Development of Nanosatellite Technology with APRS Module for Disaster Mitigation

Science.gov (United States)

Prahyang, S. Y.; Dhiya’Ulhaq, M. Z.; Golim, O. P.; Gunawan, R.; Suhandinata; Jahja, E.; Nelwan, E. R. G.; Ananta, C.; Chow, I. M.; Mali, N. D. F.

2018-05-01

Development of nanosatellite technology has enabled satellites to be developed with multiple capabilities for a specific mission in a short time with a low cost. Satellite communications are proved to be more effective in delivering information due to its large coverage area. Surya Satellite-1 will become the first Indonesian nanosatellite developed by undergraduate students. It is designed with low-cost commercial payloads, including an APRS module for communication and operated on VHF and UHF amateur radio frequencies. The mission of the satellites focused on disaster mitigation through APRS communication network with remote stations located on disaster-prone areas.

Probability Analysis of the Construction Cost of an APR1000 Single Unit

International Nuclear Information System (INIS)

Ha, Gak Hyeon; Suh, Yong Pyo; Kang, Yong Chul

2011-01-01

The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, cutting CO 2 emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the APR1000 development is to ensure export competitiveness in the developing countries in the Middle East and Southeast Asia. To that end, APR1000 (1,000MWe, 3 rd generation) will be developed based on the OPR1000 (Korean standard nuclear power plant, 2.5 generation) by incorporating and improving the general requirements such as the 60 year design life time, comprehensive site requirement of 0.3g seismic design, stability improvement, operability improvement and provisions for severe accidents. The APR1000 adds 16 advanced design features to its predecessor, as outlined below in Table 1

29 CFR 1400.735-21 - Miscellaneous statutory provisions.

Science.gov (United States)

2010-07-01

... 29 Labor 4 2010-07-01 2010-07-01 false Miscellaneous statutory provisions. 1400.735-21 Section... Service.” (b) Chapter 11 of title 18, United States Code, relating to bribery, graft, and conflicts of... Registration Act (18 U.S.C. 219). (r) Penalties: The following table, copied from the Federal Personnel Manual...

The evolution of computerized displays in accident management

International Nuclear Information System (INIS)

DeBor, J.

1988-01-01

Key regulations implemented by the NRC in 1982, which included requirements such as upgraded emergency operating procedures, detailed control room design reviews, the addition of a safety parameter display system, and the inclusion of a degreed shift technical advisor as part of the operating staff, have enabled the use of computerized displays to evolve as an integral part of accident management within each of the four main vendor groups. Problems, however, remain to be resolved in the area of technical content, information reliability, and rules for use in order to achieve the goal of more reliable accident management in nuclear power plants

Evaluation of the Main Steam Line Break Accident for the APR+ Standard Design using MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Park, M. H.; Kim, Y. S.; Hwang, Min Jeong; Sim, S. K. [Environment Energy Technology, Daejeon (Korea, Republic of); Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

As a part of licensing evaluation of the APR+ (Advanced Power Reactor +) standard design, Korea Institute of Nuclear Safety(KINS) performed safety evaluation of the APR+ Standard Safety Analysis Report(SSAR). The results of the safety evaluation of the APR+ Main Steam Line Break(MSLB) accident is presented for the most limiting post-trip return-to-power case with the single failure assumption of the Loss Of Offsite Power(LOOP). MARS-KS regulatory safety analysis code has been used to evaluate safety as well as the system behavior during MSLB accident. The MARS-KS analysis results are compared with the results of the MSLB safety analysis presented in the SSAR of the APR+. Independent safety evaluation has been performed using MARS-KS regulatory safety analysis code for the APR+ MSLB accident inside containment for the limiting case of the full power post-trip return-to-power. The results of MARS-KS analysis were compared with the results of the MSLB safety analysis presented in the APR+ SSAR. Due to higher cooldown of the MARS-KS analysis, the MARS-KS analysis results in a higher positive reactivity insertion into the core by the negative moderator and fuel temperature reactivity coefficients than the APR+ SSAR analysis. Both results show no return-to-power during the limiting case of the MSLB inside containment. However, APR+ SSAR moderator temperature reactivity insertion should be evaluated against the MARS-KS moderator density reactivity insertion for is conservatism. This study also clearly shows asymmetric thermal hydraulic behavior during the MSLB accident at intact and affected sides of the downcomer and the core. These asymmetric phenomena should be further investigated for the effects on the system design.

A Computerized Version of the Scrambled Sentences Test

Directory of Open Access Journals (Sweden)

Roberto Viviani

2018-01-01

Full Text Available The scrambled sentences test (SST, an experimental procedure that involves participants writing down their cognitions, has been used to elicit individual differences in depressiveness and vulnerability to depression. We describe here a modification of the SST to adapt it to computerized administration, with a particular view of its use in large samples and functional neuroimaging applications. In a first study with the computerized version, we reproduce the preponderance of positive cognitions in the healthy and the inverse association of these cognitions with individual measures of depressiveness. We also report a tendency of self-referential cognitions to elicit higher positive cognition rates. In a second study, we describe the patterns of neural activations elicited by emotional and neutral sentences in a functional neuroimaging study, showing that it replicates and extends previous findings obtained with the original version of the SST. During the formation of emotional cognitions, ventral areas such as the ventral anterior cingulus and the supramarginal gyrus were relatively activated. This activation pattern speaks for the recruitment of mechanisms coordinating motivational and associative processes in the formation of value-based decisions.

A study on the hydrogen distributions in a containment for nuclear plant severe accidents

Energy Technology Data Exchange (ETDEWEB)

Park, Kweon Ha; Kim, Ju Youn; Bae, Kyung Hyo [The Korea Maritime Univ., Busan (Korea, Republic of)

2012-10-15

Hydrogen explosion has been considered as one of the major issues since Fukushima nuclear accident. The cause of the explosion has not been discovered, but it is clear that the explosion strongly depends on hydrogen distributions in a containment. In this study hydrogen distributions are calculated and analyzed in the containment of APR 1400(Advanced Power Reactor 1400)

Computerized tomographic evaluation of cerebral cysticercosis

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Young; Lee, Mi Sook; Jeon, Doo Sung; Kim, Hong Soo; Rhee, Hak Song [Precbyterian Medical Center, Chonju (Korea, Republic of)

1988-08-15

Cerebral cysticercosis, unfortunately frequent in Korea, is a parastic disease in which man serve as the intermediate host of taenia solium. The larvae have a predilection for the central nervous system and can cause a variety of neurologic symptoms. The authors reviewed 19 cases of surgically proven cerebral cysticercosis and following results were obtained. 1. The most frequent age distribution was 5th and 6th decade and male to female ratio was 14:5. 2. The most frevalent involving site was cerebral parenchyme and following by ventricles. 3. Clinical manifestations were symtom and sign of increased ICP, seizure and focal neurological dificit. 4. It was assumed that computerized tomography was the procedure of choice for the diagnosis of these parasitic brain disease.

A COMPUTERIZED OPERATOR SUPPORT SYSTEM PROTOTYPE

Energy Technology Data Exchange (ETDEWEB)

Thomas A. Ulrich; Roger Lew; Ronald L. Boring; Ken Thomas

2015-03-01

A computerized operator support system (COSS) is proposed for use in nuclear power plants to assist control room operators in addressing time-critical plant upsets. A COSS is a collection of technologies to assist operators in monitoring overall plant performance and making timely, informed decisions on appropriate control actions for the projected plant condition. A prototype COSS was developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, piping and instrumentation diagram system representations, and a recommender module for mitigation actions. The initial version of the prototype is now operational at the Idaho National Laboratory using the Human System Simulation Laboratory.

Technologies in computerized lexicography | Kruyt | Lexikos

African Journals Online (AJOL)

Although the topic of this paper is technology, focus is on functional rather than technical aspects of computerized lexicography. Keywords: computerized lexicography, electronic dictionary, electronic text corpus, lexicographer's workbench, integrated language database, automatic linguistic analysis, information retrieval, ...

A feasibility assessment of calculation procedure with case study and the modification of dose conversion factor in STARDOSE

International Nuclear Information System (INIS)

Ko, H. S.; Jang, M.; Kim, S. H.; Kang, C. S.

2003-01-01

STARDOSE computer code is the designed code for which is calculated the dose of control room and off-site dose on design basis accident of nuclear power plant. Input files of STARDOSE are libfile1.txt which has decay constant and dose conversion factor, and input.dat which structurally expresses the real plant model by editor. It is given much advanced result with using newer dose conversion factor in libfile1.txt. In this study, therefore, case study is performed that is made input.dat file for LOCA and libfile1.txt which is included newer dose conversion factor and core inventory on APR-1400 and UCN 5 and 6. The result of case study is compared and analyzed

Pseudomonas syringae evades host immunity by degrading flagellin monomers with alkaline protease AprA

NARCIS (Netherlands)

Pel, Michiel J C; van Dijken, Anja J H; Bardoel, Bart W; Seidl, Michael F; van der Ent, Sjoerd; van Strijp, Jos A G; Pieterse, Corné M J

Bacterial flagellin molecules are strong inducers of innate immune responses in both mammals and plants. The opportunistic pathogen Pseudomonas aeruginosa secretes an alkaline protease called AprA that degrades flagellin monomers. Here, we show that AprA is widespread among a wide variety of

Pseudomonas syringae evades host Immunity by degrading flagellin monomers with alkaline protease AprA

NARCIS (Netherlands)

Pel, M.J.C.; Van Dijken, A.J.H.; Bardoel, B.W.; Seidl, M.F; Van der Ent, S.; Van Strijp, J.A.G.

2014-01-01

Bacterial flagellin molecules are strong inducers of innate immune responses in both mammals and plants. The opportunistic pathogen Pseudomonas aeruginosa secretes an alkaline protease called AprA that degrades flagellin monomers. Here, we show that AprA is widespread among a wide variety of

[Computerized medical record: deontology and legislation].

Science.gov (United States)

Allaert, F A; Dusserre, L

1996-02-01

Computerization of medical records is making headway for patients' follow-up, scientific research, and health expenses control, but it must not alter the guarantees provided to the patients by the medical code of ethics and the law of January 6, 1978. This law, modified on July 1, 1994, requires to register all computerized records of personal data and establishes rights to protect privacy against computer misdemeanor. All medical practitioners using computerized medical records must be aware that the infringement of this law may provoke suing in professional, civil or criminal court.

A Neural Networks Based Operation Guidance System for Procedure Presentation and Validation

International Nuclear Information System (INIS)

Seung, Kun Mo; Lee, Seung Jun; Seong, Poong Hyun

2006-01-01

In this paper, a neural network based operator support system is proposed to reduce operator's errors in abnormal situations in nuclear power plants (NPPs). There are many complicated situations, in which regular and suitable operations should be done by operators accordingly. In order to regulate and validate operators' operations, it is necessary to develop an operator support system which includes computer based procedures with the functions for operation validation. Many computerized procedures systems (CPS) have been recently developed. Focusing on the human machine interface (HMI) design and procedures' computerization, most of CPSs used various methodologies to enhance system's convenience, reliability and accessibility. Other than only showing procedures, the proposed system integrates a simple CPS and an operation validation system (OVS) by using artificial neural network (ANN) for operational permission and quantitative evaluation

39 CFR 501.15 - Computerized Meter Resetting System.

Science.gov (United States)

2010-07-01

... 39 Postal Service 1 2010-07-01 2010-07-01 false Computerized Meter Resetting System. 501.15... AND DISTRIBUTE POSTAGE EVIDENCING SYSTEMS § 501.15 Computerized Meter Resetting System. (a) Description. The Computerized Meter Resetting System (CMRS) permits customers to reset their postage meters at...

The Use of Statistical Process Control-Charts for Person-Fit Analysis on Computerized Adaptive Testing. LSAC Research Report Series.

Science.gov (United States)

Meijer, Rob R.; van Krimpen-Stoop, Edith M. L. A.

In this study a cumulative-sum (CUSUM) procedure from the theory of Statistical Process Control was modified and applied in the context of person-fit analysis in a computerized adaptive testing (CAT) environment. Six person-fit statistics were proposed using the CUSUM procedure, and three of them could be used to investigate the CAT in online test…

NCBI nr-aa BLAST: CBRC-OPRI-01-1400 [SEVENS

Lifescience Database Archive (English)

Full Text Available ck Protein 1a (Hsp70-1) Atpase Domain In Complex With Adp And Inorganic Phosphate 3JXU 1e-171 94% ... ...CBRC-OPRI-01-1400 pdb|3JXU|A Chain A, Crystal Structure Of The Human 70kda Heat Sho

APR-246/PRIMA-1Met Inhibits and Reverses Squamous Metaplasia in Human Conjunctival Epithelium.

Science.gov (United States)

Li, Jing; Li, Cheng; Wang, Guoliang; Liu, Zhen; Chen, Pei; Yang, Qichen; Dong, Nuo; Wu, Huping; Liu, Zuguo; Li, Wei

2016-02-01

Squamous metaplasia is a common pathologic condition in ocular surface diseases for which there is no therapeutic medication in clinic. In this study, we investigated the effect of a small molecule, APR-246/PRIMA-1(Met), on squamous metaplasia in human conjunctival epithelium. Human conjunctival explants were cultured for up to 12 days under airlifting conditions. Epithelial cell differentiation and proliferation were assessed by Cytokeratin 10 (K10), K14, K19, Pax6, MUC5AC, and p63 immunostaining patterns. β-catenin and TCF-4 immunofluorescent staining and real-time PCR characterized Wnt signaling pathway involvement. Pterygium clinical samples were cultured under airlifting conditions with or without APR-246 for 4 days. p63, K10, β-catenin, and TCF-4 expression in pterygial epithelium was determined by immunofluorescent staining and real-time PCR. Airlift conjunctival explants resulted in increased stratification and intrastromal epithelial invagination. Such pathology was accompanied by increases in K10, K14, and p63 expression, whereas K19 and Pax6 levels declined when compared to those in freshly isolated tissue. On the other hand, APR-246 reversed all of these declines in K10, K14, and p63 expression. Furthermore, K19 and Pax6 increased along with rises in goblet cell density. These effects of APR-246 were accompanied by near restoration of normal conjunctival epithelial histology. APR-246 also reversed squamous metaplasia in pterygial epithelium that had developed after 4 days in ex vivo culture. Reductions in squamous metaplasia induced by APR-246 suggest it may provide a novel therapeutic approach in different squamous metaplasia-associated ocular surface diseases.

Determining women's sexual self-schemas through advanced computerized text analysis.

Science.gov (United States)

Stanton, Amelia M; Boyd, Ryan L; Pulverman, Carey S; Meston, Cindy M

2015-08-01

The meaning extraction method (MEM), an advanced computerized text analysis technique, was used to analyze women's sexual self-schemas. Participants (n=239) completed open-ended essays about their personal feelings associated with sex and sexuality. These essays were analyzed using the MEM, a procedure designed to extract common themes from natural language. Using the MEM procedure, we extracted seven unique themes germane to sexual self-schemas: family and development, virginity, abuse, relationship, sexual activity, attraction, and existentialism. Each of these themes is comprised of frequently used words across the participants' descriptions of their sexual selves. Significant differences in sexual self-schemas were observed to covary with age, relationship status, and sexual abuse history. Copyright © 2015 Elsevier Ltd. All rights reserved.

Strategies for Controlling Item Exposure in Computerized Adaptive Testing with the Generalized Partial Credit Model

Science.gov (United States)

Davis, Laurie Laughlin

2004-01-01

Choosing a strategy for controlling item exposure has become an integral part of test development for computerized adaptive testing (CAT). This study investigated the performance of six procedures for controlling item exposure in a series of simulated CATs under the generalized partial credit model. In addition to a no-exposure control baseline…

Microcomputer Network for Computerized Adaptive Testing (CAT)

Science.gov (United States)

1984-03-01

PRDC TR 84-33 \\Q.�d-33- \\ MICROCOMPUTER NETWOJlt FOR COMPUTERIZED ADAPTIVE TESTING ( CAT ) Baldwin Quan Thomas A . Park Gary Sandahl John H...ACCEIIION NO NPRDC TR 84-33 4. TITLE (-d Sul>tlllo) MICROCOMP UTER NETWORK FOR COMPUTERIZED ADA PTIVE TESTING ( CAT ) 1. Q B. uan T. A . Park...adaptive testing ( CAT ) Bayesian sequential testing 20. ABSTitACT (Continuo on ro•••• aide II noco .. _, _., ld-tlly ,.,. t.loclt _._.) DO Computerized

Retort Racks for Polymeric Trays in 1400 Style Spray Retorts

National Research Council Canada - National Science Library

Bruins, Henderikus B

2003-01-01

The objective of this project was to design a retort rack that would maximize the retort capacity of a 1400 style spray retort and to select a material that would withstand the harsh retort spray environment...

Determining Women’s Sexual Self-Schemas Through Advanced Computerized Text Analysis

Science.gov (United States)

Stanton, Amelia M.; Boyd, Ryan L.; Pulverman, Carey S.; Meston, Cindy M.

2015-01-01

The meaning extraction method (MEM), an advanced computerized text analysis technique, was used to analyze women’s sexual self-schemas. Participants (n = 239) completed open-ended essays about their personal feelings associated with sex and sexuality. These essays were analyzed using the MEM, a procedure designed to extract common themes from natural language. Using the MEM procedure, we extracted seven unique themes germane to sexual self-schemas: family and development, virginity, abuse, relationship, sexual activity, attraction, and existentialism. Each of these themes is comprised of frequently used words across the participants’ descriptions of their sexual selves. Significant differences in sexual self-schemas were observed to covary with age, relationship status, and sexual abuse history. PMID:26146161

Computerized system for measuring cerebral metabolism

International Nuclear Information System (INIS)

McGlone, J.S.; Hibbard, L.S.; Hawkins, R.A.; Kasturi, R.

1987-01-01

A computerized stereotactic measurement system for evaluating rat brain metabolism was developed to utilize the large amount of data generated by quantitative autoradiography. Conventional methods of measurement only analyze a small percent of these data because these methods are limited by instrument design and the subjectiveness of the investigator. However, a computerized system allows digital images to be analyzed by placing data at their appropriate three-dimensional stereotactic coordinates. The system automatically registers experimental data to a standard three-dimensional image using alignment, scaling, and matching operations. Metabolic activity in different neuronal structures is then measured by generating digital masks and superimposing them on to experimental data. Several experimental data sets were evaluated and it was noticed that the structures measured by the computerized system, had in general, lower metabolic activity than manual measurements had indicated. This was expected because the computerized system measured the structure over its volume while the manual readings were taken from the most active metabolic area of a particular structure

Software Enhancement of SKN 3 and 4 Tagout Management System for MCR operators

International Nuclear Information System (INIS)

Lee, Sungjin; Seong, Nokyu; Jung Yeonsub

2014-01-01

One of the human performance tools for preventing human errors is tagging and operational barrier. Tagging is useful for checking correct equipment status during and after maintenance and test activities. A tag is systematically generated by related workflow procedure and Enterprise Resource Planning (ERP) in most nuclear power plant of KHNP. With the introduction of Advanced Power Reactor 1400 (APR1400) Man-Machine Interface System (MMIS), Main Control Room (MCR) operators' tagging is performed in specific software application such as 'Tagging Administration'. This paper suggests enhanced user interface and work flow of the tagout management system. Enhanced TMS for SKN 3 and 4 can be useful tools for operators' tagging activity. More end-users needs can be founded until the completion of enhancement. Enhanced TMS is independently working with IPS so that it can be applied into other nuclear power plant which is constructing or operating

Software Enhancement of SKN 3 and 4 Tagout Management System for MCR operators

Energy Technology Data Exchange (ETDEWEB)

Lee, Sungjin; Seong, Nokyu; Jung Yeonsub [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2014-05-15

One of the human performance tools for preventing human errors is tagging and operational barrier. Tagging is useful for checking correct equipment status during and after maintenance and test activities. A tag is systematically generated by related workflow procedure and Enterprise Resource Planning (ERP) in most nuclear power plant of KHNP. With the introduction of Advanced Power Reactor 1400 (APR1400) Man-Machine Interface System (MMIS), Main Control Room (MCR) operators' tagging is performed in specific software application such as 'Tagging Administration'. This paper suggests enhanced user interface and work flow of the tagout management system. Enhanced TMS for SKN 3 and 4 can be useful tools for operators' tagging activity. More end-users needs can be founded until the completion of enhancement. Enhanced TMS is independently working with IPS so that it can be applied into other nuclear power plant which is constructing or operating.

How to Train an Injured Brain? A Pilot Feasibility Study of Home-Based Computerized Cognitive Training.

Science.gov (United States)

Verhelst, Helena; Vander Linden, Catharine; Vingerhoets, Guy; Caeyenberghs, Karen

2017-02-01

Computerized cognitive training programs have previously shown to be effective in improving cognitive abilities in patients suffering from traumatic brain injury (TBI). These studies often focused on a single cognitive function or required expensive hardware, making it difficult to be used in a home-based environment. This pilot feasibility study aimed to evaluate the feasibility of a newly developed, home-based, computerized cognitive training program for adolescents who suffered from TBI. Additionally, feasibility of study design, procedures, and measurements were examined. Case series, longitudinal, pilot, feasibility intervention study with one baseline and two follow-up assessments. Nine feasibility outcome measures and criteria for success were defined, including accessibility, training motivation/user experience, technical smoothness, training compliance, participation willingness, participation rates, loss to follow-up, assessment timescale, and assessment procedures. Five adolescent patients (four boys, mean age = 16 years 7 months, standard deviation = 9 months) with moderate to severe TBI in the chronic stage were recruited and received 8 weeks of cognitive training with BrainGames. Effect sizes (Cohen's d) were calculated to determine possible training-related effects. The new cognitive training intervention, BrainGames, and study design and procedures proved to be feasible; all nine feasibility outcome criteria were met during this pilot feasibility study. Estimates of effect sizes showed small to very large effects on cognitive measures and questionnaires, which were retained after 6 months. Our pilot study shows that a longitudinal intervention study comprising our novel, computerized cognitive training program and two follow-up assessments is feasible in adolescents suffering from TBI in the chronic stage. Future studies with larger sample sizes will evaluate training-related effects on cognitive functions and underlying brain structures.

ATLAS Facility and Instrumentation Description Report

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Moon, Sang Ki; Park, Hyun Sik

2009-06-01

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). The ATLAS is a half-height and 1/288-volume scaled test facility with respect to the APR1400. The fluid system of the ATLAS consists of a primary system, a secondary system, a safety injection system, a break simulating system, a containment simulating system, and auxiliary systems. The primary system includes a reactor vessel, two hot legs, four cold legs, a pressurizer, four reactor coolant pumps, and two steam generators. The secondary system of the ATLAS is simplified to be of a circulating looptype. Most of the safety injection features of the APR1400 and the OPR1000 are incorporated into the safety injection system of the ATLAS. In the ATLAS test facility, about 1300 instrumentations are installed to precisely investigate the thermal-hydraulic behavior in simulation of the various test scenarios. This report describes the scaling methodology, the geometric data of the individual component, and the specification and the location of the instrumentations which are specific to the simulation of 50% DVI line break accident of the APR1400 for supporting the 50 th OECD/NEA International Standard Problem Exercise (ISP-50)

Evaluation of brain scintigraphy and computerized tomography

International Nuclear Information System (INIS)

Cavailloles, F.; Dairou, R.; Desbleds, M.T.; Benoit, C.; Larmande, P.; Bok, B.; Alperovitch, A.

1983-01-01

In order to assess the clinical usefulness of brain computerized tomography and radionuclide scan, a prospective study was performed on a series of 554 patients. The detection rate was assessed as well as the identification rate of lesions. In addition, the usefulness of both tests was appreciated subjectively by two neurologists reviewing the patients' files. Both give reasonably similar results: computerized tomography is superior to radionuclide scan in the diagnosis of tumors and intracerebral hematomas, the radionuclide scan being slightly superior in the diagnosis of purely ischemic CVA and subdural hematomas. The superiority which was subjectively conceded to computerized tomography was greater than that objectively demonstrated. However, clinical usefulness of computerized tomography was judged important in only 50% of the cases. Moreover, to request both computerized tomography and radionuclide scan appeared as having no interest in 83% of the cases. In that series, the diagnostic hypotheses were in agreement with the final diagnosis in 88% of the cases. Bias encountered in this type of studies are discussed [fr

Computerized tomography in myotonic dystrophy

International Nuclear Information System (INIS)

Gellerich, I.; Mueller, D.; Koch, R.D.

1986-01-01

Besides clinical symptoms, progress and electromyography computerized tomography improves the diagnostics of myotonic dystrophy. Even small changes in muscular structure are detectable and especially the musculus soleus exhibits early and pronounced alterations. By means of density distribution pattern an improved characterization of the disease is possible. Additional information is obtained by cerebral computerized tomography. Atrophy of brain tissue is to be expected in all patients with myotonic dystrophy. (author)

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

The Ongoing Rediscovery of Après-Coup as a Central Freudian Concept.

Science.gov (United States)

House, Jonathan

2017-10-01

Après-coup, Freud's Nachträglichkeit, is an essential psychoanalytic concept structuring each of four concepts, four mental processes that lie at the foundation of Freud's thinking: psychic trauma, repression, the creation of the unconscious, and the creation of infantile sexuality. It is argued here that infantile sexual drives, in contrast to the self-preservative instincts, arise from a two-step process of translation and repression in which the residues of failed translation become source-objects of the drives. These residues of failed translation have an associative resonance with adult sexuality, and the child is driven to ongoing attempts to translate them, to make them meaningful après coup. Thus, après-coup is at the heart of the human subject as a sexual creature who requires, desires, and creates meaning.

Deletion of aprA and nprA genes for alkaline protease A and neutral protease A from bacillus thuringiensis: effect on insecticidal crystal proteins.

Science.gov (United States)

Tan, Y; Donovan, W P

2001-11-17

The aprA gene encoding alkaline protease A (AprA) was cloned from Bacillus thuringiensis subsp. kurstaki, and the cloned gene was used to construct aprA-deleted (aprA1) strains of B. thuringiensis. An aprA1 strain of B. thuringiensis that contained the wild-type gene for neutral protease A (nprA(+)) displayed levels of extracellular proteolytic activity that were similar to those of an aprA(+)nprA(+) strain. However, when EDTA was included in the protease assay to inhibit NprA activity the aprA1nprA(+) strain displayed only 2% of the extracellular proteolytic activity of the aprA(+)nprA(+) strain. A strain that was deleted for both aprA and nprA (aprA1nprA3 strain) failed to produce detectable levels of proteolytic activity either in the presence or absence of EDTA in the assay. Compared with the aprA(+)nprA(+) strain the aprA1nprA(+) strain yielded 10% more full-length Cry1Bb crystal protein and the aprA1nprA3 strain yielded 25% more full-length Cry1Bb protein. No significant differences were seen in the 50% lethal dose of Cry1Bb protein from aprA(+)nprA(+) and aprA1nprA3 strains against three species of lepidopteran insects. These results suggest that enhanced yield of certain crystal proteins can be obtained by deletion of the genes aprA and nprA which are the major extracellular proteases of B. thuringiensis.

Homology modeling of dissimilatory APS reductases (AprBA of sulfur-oxidizing and sulfate-reducing prokaryotes.

Directory of Open Access Journals (Sweden)

Birte Meyer

Full Text Available BACKGROUND: The dissimilatory adenosine-5'-phosphosulfate (APS reductase (cofactors flavin adenine dinucleotide, FAD, and two [4Fe-4S] centers catalyzes the transformation of APS to sulfite and AMP in sulfate-reducing prokaryotes (SRP; in sulfur-oxidizing bacteria (SOB it has been suggested to operate in the reverse direction. Recently, the three-dimensional structure of the Archaeoglobus fulgidus enzyme has been determined in different catalytically relevant states providing insights into its reaction cycle. METHODOLOGY/PRINCIPAL FINDINGS: Full-length AprBA sequences from 20 phylogenetically distinct SRP and SOB species were used for homology modeling. In general, the average accuracy of the calculated models was sufficiently good to allow a structural and functional comparison between the beta- and alpha-subunit structures (78.8-99.3% and 89.5-96.8% of the AprB and AprA main chain atoms, respectively, had root mean square deviations below 1 A with respect to the template structures. Besides their overall conformity, the SRP- and SOB-derived models revealed the existence of individual adaptations at the electron-transferring AprB protein surface presumably resulting from docking to different electron donor/acceptor proteins. These structural alterations correlated with the protein phylogeny (three major phylogenetic lineages: (1 SRP including LGT-affected Archaeoglobi and SOB of Apr lineage II, (2 crenarchaeal SRP Caldivirga and Pyrobaculum, and (3 SOB of the distinct Apr lineage I and the presence of potential APS reductase-interacting redox complexes. The almost identical protein matrices surrounding both [4Fe-4S] clusters, the FAD cofactor, the active site channel and center within the AprB/A models of SRP and SOB point to a highly similar catalytic process of APS reduction/sulfite oxidation independent of the metabolism type the APS reductase is involved in and the species it has been originated from. CONCLUSIONS: Based on the comparative

Anti-3-[(18)F]FACBC positron emission tomography-computerized tomography and (111)In-capromab pendetide single photon emission computerized tomography-computerized tomography for recurrent prostate carcinoma: results of a prospective clinical trial.

Science.gov (United States)

Schuster, David M; Nieh, Peter T; Jani, Ashesh B; Amzat, Rianot; Bowman, F Dubois; Halkar, Raghuveer K; Master, Viraj A; Nye, Jonathon A; Odewole, Oluwaseun A; Osunkoya, Adeboye O; Savir-Baruch, Bital; Alaei-Taleghani, Pooneh; Goodman, Mark M

2014-05-01

We prospectively evaluated the amino acid analogue positron emission tomography radiotracer anti-3-[(18)F]FACBC compared to ProstaScint® ((111)In-capromab pendetide) single photon emission computerized tomography-computerized tomography to detect recurrent prostate carcinoma. A total of 93 patients met study inclusion criteria who underwent anti-3-[(18)F]FACBC positron emission tomography-computerized tomography plus (111)In-capromab pendetide single photon emission computerized tomography-computerized tomography for suspected recurrent prostate carcinoma within 90 days. Reference standards were applied by a multidisciplinary board. We calculated diagnostic performance for detecting disease. In the 91 of 93 patients with sufficient data for a consensus on the presence or absence of prostate/bed disease anti-3-[(18)F]FACBC had 90.2% sensitivity, 40.0% specificity, 73.6% accuracy, 75.3% positive predictive value and 66.7% negative predictive value compared to (111)In-capromab pendetide with 67.2%, 56.7%, 63.7%, 75.9% and 45.9%, respectively. In the 70 of 93 patients with a consensus on the presence or absence of extraprostatic disease anti-3-[(18)F]FACBC had 55.0% sensitivity, 96.7% specificity, 72.9% accuracy, 95.7% positive predictive value and 61.7% negative predictive value compared to (111)In-capromab pendetide with 10.0%, 86.7%, 42.9%, 50.0% and 41.9%, respectively. Of 77 index lesions used to prove positivity histological proof was obtained in 74 (96.1%). Anti-3-[(18)F]FACBC identified 14 more positive prostate bed recurrences (55 vs 41) and 18 more patients with extraprostatic involvement (22 vs 4). Anti-3-[(18)F]FACBC positron emission tomography-computerized tomography correctly up-staged 18 of 70 cases (25.7%) in which there was a consensus on the presence or absence of extraprostatic involvement. Better diagnostic performance was noted for anti-3-[(18)F]FACBC positron emission tomography-computerized tomography than for (111)In-capromab pendetide single

Constrained model predictive control for load-following operation of APR reactors

International Nuclear Information System (INIS)

Kim, Jae Hwan; Lee, Sim Won; Kim, Ju Hyun; Na, Man Gyun; Yu, Keuk Jong; Kim, Han Gon

2012-01-01

The load-following operation of APR+ reactor is needed to control the power effectively using the control rods and to restrain the reactivity control from using the boric acid for flexibility of plant operation. Usually, the reason why the disproportion of axial flux distribution occurs during load-following operation is xenon-induced oscillation. The xenon has a very high absorption cross-section and makes the impact on the reactor delayed by the iodine precursor. The power maneuvering using automatically load-following operation has advantage in terms of safety and economic operation of the reactor, so the controller has to be designed efficiently. Therefore, an advanced control method that meets the conditions such as automatic control, flexibility, safety, and convenience is necessary to load-following operation of APR+ reactor. In this paper, the constrained model predictive control (MPC) method is applied to design APR reactor's automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control. Some controllers use only the current tracking command, but MPC considers future commands in addition to the current tracking command. So, MPC can achieve better tracking performance than others. Furthermore, an MPC is to used in many industrial process control systems. The basic concept of the MPC is to solve an optimization problem for a finite future time interval at present time and to implement the first optimal control input as the current control input. The KISPAC-1D code, which models the APR+ nuclear power plants, is interfaced to the proposed controller to verify the tracking performance of the reactor power level and ASI. It is known that the proposed controller exhibits very fast tracking responses

NCBI nr-aa BLAST: CBRC-SARA-01-1400 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-SARA-01-1400 gb|AAW70053.1| MRGX2 [Homo sapiens] gb|AAW70054.1| MRGX2 [Homo sapiens...] gb|AAW70055.1| MRGX2 [Homo sapiens] gb|AAW70070.1| MRGX2 [Homo sapiens] gb|AAW70083.1| MRGX2 [Homo sapiens] AAW70053.1 4e-34 52% ...

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Guide to data processing and revision: Part 3, Hardware component failure data entry and revision procedures

International Nuclear Information System (INIS)

Gilmore, W.E.; Gertman, D.I.; Gilbert, B.G.; Reece, W.J.

1988-11-01

The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability (HEP) and hardware component failure data (HCFD). The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. Users can perform data base searches to furnish HEP estimates and HCFD rates. In this manner, the NUCLARR system can be used to support a variety of risk assessment activities. This volume, Volume 3 of a 5-volume series, presents the procedures used to process HEP and HCFD for entry in NUCLARR and describes how to modify the existing NUCLARR taxonomy in order to add equipment types of action verbs. Volume 3 also specifies the various roles of the administrative staff on assignment to the NUCLARR Clearinghouse who are tasked with maintaining the data base, dealing with user requests, and processing NUCLARR data

Vingt ans après…

Directory of Open Access Journals (Sweden)

François Burgat

2002-05-01

Full Text Available Le dixième numéro des Chroniques Yéménites, publié vingt ans après la création du Centre Français d'Archéologie et de Sciences Sociales de Sanaa reflète des ambitions souvent exprimées: donner une chance aux jeunes chercheurs et couvrir un large éventail disciplinaire au service de l'affirmation de la vocation régionale du Centre. Depuis Sanaa ou Paris, chercheurs, doctorants ou stagiaires, de l'Ecole normale supérieure, de l'INALCO, de l'Université de Sanaa et de bon nombre d'institu...

Computerized engineering logic for procurement and dedication processes

International Nuclear Information System (INIS)

Tulay, M.P.

1996-01-01

This paper summarizes the work performed for designing the system and especially for calculating on-line expected performance and gives some significant results. In an attempt to better meet the needs of operations and maintenance organizations, many nuclear utility procurement engineering groups have simplified their procedures, developed on-line tools for performing the specification of replacement items, and developed relational databases containing part-level information necessary to automate the procurement process. Although these improvements have helped to reduce the engineering necessary to properly specify and accept/dedicate items for nuclear safety-related applications, a number of utilities have recognized that additional long-term savings can be realized by integrating a computerized logic to assist technical procurement engineering personnel

77 FR 15577 - Standard Instrument Approach Procedures, and Takeoff Minimums and Obstacle Departure Procedures...

Science.gov (United States)

2012-03-16

...-Apr-12 UT Logan Logan-Cache..... 2/3187 2/23/12 RNAV (GPS) RWY 17, Amdt 1. 5-Apr-12 VT Lyndonville... Paul Intl/Wold- 30L, ILS RWY Chamberlain. 30L (CAT II), Amdt 44A. 5-Apr-12 MN Minneapolis..... Minneapolis-St 2/6419 2/23/12 CONVERGING ILS Paul Intl/Wold- RWY 30L, Orig. Chamberlain. 5-Apr-12 MN Duluth...

aprABC: A Mycobacterium tuberculosis complex-specific locus that modulates pH-driven adaptation to the macrophage phagosome

Science.gov (United States)

Abramovitch, Robert B.; Rohde, Kyle H.; Hsu, Fong-Fu; Russell, David G.

2011-01-01

Summary Following phagocytosis by macrophages, Mycobacterium tuberculosis (Mtb) senses the intracellular environment and remodels its gene expression for growth in the phagosome. We have identified an Acid and Phagosome Regulated (aprABC) locus that is unique to the Mtb complex and whose gene expression is induced during growth in acidic environments in vitro and in macrophages. Using the aprA promoter, we generated a strain that exhibits high levels of inducible fluorescence in response to growth in acidic medium in vitro and in macrophages. aprABC expression is dependent on the two-component regulator phoPR, linking phoPR signaling to pH sensing. Deletion of the aprABC locus causes defects in gene expression that impact aggregation, intracellular growth, and the relative levels of storage and cell wall lipids. We propose a model where phoPR senses the acidic pH of the phagosome and induces aprABC expression to fine-tune processes unique for intracellular adaptation of Mtb complex bacteria. PMID:21401735

Resultats visuels et complications apres chirurgie de la cataracte ...

African Journals Online (AJOL)

Resultats visuels et complications apres chirurgie de la cataracte: cas de l' Hopital Ophtalmologique Saint Andre de Tinre au nord-Benin. K.M. Amedome, C.R.A. Assavedo, M Aboudou, K Vonor, N Maneh, K Nonon Saa, K Dzidzinyo, K.D. Ayena, M Banla, K Balo ...

Computerized Classification Testing with the Rasch Model

Science.gov (United States)

Eggen, Theo J. H. M.

2011-01-01

If classification in a limited number of categories is the purpose of testing, computerized adaptive tests (CATs) with algorithms based on sequential statistical testing perform better than estimation-based CATs (e.g., Eggen & Straetmans, 2000). In these computerized classification tests (CCTs), the Sequential Probability Ratio Test (SPRT) (Wald,…

Evaluation of the Automatic Density Compensation for Pressurizer Level Measurement

International Nuclear Information System (INIS)

Jeong, Insoo; Min, Seohong; Ahn, Myunghoon

2014-01-01

When using two transmitters, it is difficult for the operators to identify the correct level of the pressurizer (PZR) upon failure of one of the two transmitters. For this reason, Korean Utility Requirements Document (KURD) requires that the operators to use three independent level indicators. Two hot calibrated transmitters and one cold calibrated transmitter compose PZR level transmitters in APR1400. In this paper, the deviation between cold calibration and hot calibration is evaluated, and the application of compensated PZR level measurement and uncompen-sated PZR level measurement during the normal operation of APR1400 are introduced. The PZR level signals for APR1400 come in three channels. To satisfy the KURD requirements for PZR level measurement, and at the same time to accomplish correction design and implementation, applicability and differences between hot calibration and cold calibration, compensated level and uncompensated level were evaluated as follows: For proper indication of PZR levels under normal operating condition, two of the three transmitters went through hot calibration and the remaining one transmitter went through cold calibration. This was to allow indicating entire regions of PZR regardless of the plant operation modes. For automatic density compensation per KURD requirements, the algorithm of the density compensated PZR level implemented in the DCS controller and PRV logic is adopted as a signal validation method

ATLAS Facility Description Report

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Moon, Sang Ki; Park, Hyun Sik; Cho, Seok; Choi, Ki Yong

2009-04-01

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). The ATLAS has the same two-loop features as the APR1400 and is designed according to the well-known scaling method suggested by Ishii and Kataoka to simulate the various test scenarios as realistically as possible. It is a half-height and 1/288-volume scaled test facility with respect to the APR1400. The fluid system of the ATLAS consists of a primary system, a secondary system, a safety injection system, a break simulating system, a containment simulating system, and auxiliary systems. The primary system includes a reactor vessel, two hot legs, four cold legs, a pressurizer, four reactor coolant pumps, and two steam generators. The secondary system of the ATLAS is simplified to be of a circulating loop-type. Most of the safety injection features of the APR1400 and the OPR1000 are incorporated into the safety injection system of the ATLAS. In the ATLAS test facility, about 1300 instrumentations are installed to precisely investigate the thermal-hydraulic behavior in simulation of the various test scenarios. This report describes the scaling methodology, the geometric data of the individual component, and the specification and the location of the instrumentations in detail

Double-Ended Break Test of an 8.5 inch Direct Vessel Injection Line using the ATLAS

International Nuclear Information System (INIS)

Choi, Ki Yong; Kang, Kyoung Ho; Kim, Bok Deuk; Kim, Yeon Sik; Min, Kyoung Ho; Park, Choon Kyoung; Park, Hyun Sik; Baek, Won Pil; Cho, Seok; Choi, Nam Hyun

2010-01-01

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at the Korea Atomic Energy Research Institute (KAERI). It is a 1/2 reduced height and 1/288-volume scaled test facility with respect to the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. In 2008, an integral effect test for simulating a guillotine break of a DVI line of the APR1400 was carried out as the first DVI test item, named as SB-DVI-03, on May, 2008. With an improvement on the break flow measuring system, the second DVI test for a guillotine break of a DVI line of the APR1400, named as SB-DVI-08, was conducted for repeatability. The present data is the first integral effect test data of its kind for simulating a DVI line break accident. It will help in understanding the thermal hydraulic phenomena occurring during the DVI line break accident. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. The present integral effect test data will be used to validate the current safety analysis methodology for the DVI line break accident

Single Tube Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of a single U-tube in the APR1400, the SGTR-CL-01 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Following the reactor trip induced by high steam generator level (HSGL) signal, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened. The MSSVs repeated on and off status depending on the secondary system pressure during the whole test period. Due to the break flow, the collapsed water level of the affected steam generator showed milder decrease than that of the intact steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for a SGTR simulation, especially for DVI-adapted plants

Conceptual Design of Condenser using PCM as PCCS for APR1400

International Nuclear Information System (INIS)

Jung, Hwa-Young; Ko, A-Reum; Lee, Jeong Ik; Yoon, Ho Joon

2016-01-01

In this study, the conceptual design of the condenser using the phase change material (PCM) is proposed as another PCCS concept for PWRs. It is possible to install the condenser for the existing nuclear power plants under operation without PCCS. Most of all, this condenser has to be designed to reduce the pressure and temperature in the containment until the containment spray system is available. A feature of the condenser as shown in Fig. 2 is the composition of PCM and wire mesh connected with the fins to improve heat transfer. As the first step, a preliminary study was conducted to find the optimum design of the part at which the condensation occurs under the given conditions. The results show that the fin base area as well as the total heat transfer area of fins are generally in inversely proportional to the average condensation heat transfer coefficient. Also, the point at which the gradient is changed is when the flow regime is changed from wavy flow to turbulent flow. However, the number of fins shows the different trend. Thus, in order to achieve better average condensation heat transfer coefficient under the identical heat capacity condition, the shorter the fin length and the more fins are necessary for the optimum design the condenser in wavy flow region

Conceptual Design of Condenser using PCM as PCCS for APR1400

Energy Technology Data Exchange (ETDEWEB)

Jung, Hwa-Young; Ko, A-Reum; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-10-15

In this study, the conceptual design of the condenser using the phase change material (PCM) is proposed as another PCCS concept for PWRs. It is possible to install the condenser for the existing nuclear power plants under operation without PCCS. Most of all, this condenser has to be designed to reduce the pressure and temperature in the containment until the containment spray system is available. A feature of the condenser as shown in Fig. 2 is the composition of PCM and wire mesh connected with the fins to improve heat transfer. As the first step, a preliminary study was conducted to find the optimum design of the part at which the condensation occurs under the given conditions. The results show that the fin base area as well as the total heat transfer area of fins are generally in inversely proportional to the average condensation heat transfer coefficient. Also, the point at which the gradient is changed is when the flow regime is changed from wavy flow to turbulent flow. However, the number of fins shows the different trend. Thus, in order to achieve better average condensation heat transfer coefficient under the identical heat capacity condition, the shorter the fin length and the more fins are necessary for the optimum design the condenser in wavy flow region.

Turbulent convection experiment at high Rayleigh number to support CAP1400 IVR strategy

Energy Technology Data Exchange (ETDEWEB)

Ma, Li, E-mail: mali@snptrd.com [State Nuclear Hua Qing(Beijing) Nuclear Power Technology R& D Centre Co., Ltd, Building A, State Nuclear Power Research Institute, Future Science & Technology Park, Changping Dist., Beijing 102209 (China); Li, Jing, E-mail: lijing@snptrd.com [State Nuclear Hua Qing(Beijing) Nuclear Power Technology R& D Centre Co., Ltd, Building A, State Nuclear Power Research Institute, Future Science & Technology Park, Changping Dist., Beijing 102209 (China); Ji, Shui, E-mail: jishui@snptrd.com [State Nuclear Hua Qing(Beijing) Nuclear Power Technology R& D Centre Co., Ltd, Building A, State Nuclear Power Research Institute, Future Science & Technology Park, Changping Dist., Beijing 102209 (China); Chang, Huajian, E-mail: changhuajian@snptrd.com [State Nuclear Hua Qing(Beijing) Nuclear Power Technology R& D Centre Co., Ltd, Building A, State Nuclear Power Research Institute, Future Science & Technology Park, Changping Dist., Beijing 102209 (China); Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

2015-10-15

Highlights: • The facility reached high Ra number at 10{sup 12} of CAP1400 working condition. • The fitting formula Nu = 0.085 × Ra{sup 0.315} was established to calculate the heat flux in the metal layer at high Ra for the CAP1400. • The coupling method can accurately and safely predict the heat flow distribution of metal layer in high Ra number conditions. • The experiment results will predict the relationship between axial and radial heat transfer well. - Abstract: The characteristics of the heat transfer and the calculation of heat flux in metal layer are both the critical problems for in-vessel retention (IVR) strategy. Turbulent convection occurs in the metal layer when the Rayleigh number (Ra) becomes sufficient high. The Globe–Dropkin (G–D) correlation (Globe and Dropkin, 1959) and Chu–Churchill (C–C) correlation (Churchill and Chu, 1975) have been widely used to calculate the heat flux in the metal layer, where the valid range of the Ra is from 1.5 × 10{sup 5} to 6.8 × 10{sup 8} in G–D correlation and less than 10{sup 12} in C–C correlation. However, with the increase of reactor power, both the Rayleigh number and the rate of heat transfer below the bottom of metal layer of the molten pool will increase, and in this case the Rayleigh number even can reach 10{sup 11} for the China Advanced Passive Plant CAP1400. Accordingly, the G–D correlation is not suitable for the CAP1400. Therefore, our experiment purposes are to establish the appropriate correlation at high Ra for the CAP1400 and predict the axial and radial distribution of the heat transfer in the metal layer with the heat transfer behavior of metal layer experiment (HELM) facility. The experiments are divided into two parts. Each part concerns 39 runs and 47 experimental conditions. Its corresponding results are obtained at middle Prandtl number (Pr = 7 for water) and the Nusselt number is found to be proportional to Ra{sup 0.315} in the range 3.93 × 10{sup 8} < Ra < 3.57 �

Turbulent convection experiment at high Rayleigh number to support CAP1400 IVR strategy

International Nuclear Information System (INIS)

Ma, Li; Li, Jing; Ji, Shui; Chang, Huajian

2015-01-01

Highlights: • The facility reached high Ra number at 10 12 of CAP1400 working condition. • The fitting formula Nu = 0.085 × Ra 0.315 was established to calculate the heat flux in the metal layer at high Ra for the CAP1400. • The coupling method can accurately and safely predict the heat flow distribution of metal layer in high Ra number conditions. • The experiment results will predict the relationship between axial and radial heat transfer well. - Abstract: The characteristics of the heat transfer and the calculation of heat flux in metal layer are both the critical problems for in-vessel retention (IVR) strategy. Turbulent convection occurs in the metal layer when the Rayleigh number (Ra) becomes sufficient high. The Globe–Dropkin (G–D) correlation (Globe and Dropkin, 1959) and Chu–Churchill (C–C) correlation (Churchill and Chu, 1975) have been widely used to calculate the heat flux in the metal layer, where the valid range of the Ra is from 1.5 × 10 5 to 6.8 × 10 8 in G–D correlation and less than 10 12 in C–C correlation. However, with the increase of reactor power, both the Rayleigh number and the rate of heat transfer below the bottom of metal layer of the molten pool will increase, and in this case the Rayleigh number even can reach 10 11 for the China Advanced Passive Plant CAP1400. Accordingly, the G–D correlation is not suitable for the CAP1400. Therefore, our experiment purposes are to establish the appropriate correlation at high Ra for the CAP1400 and predict the axial and radial distribution of the heat transfer in the metal layer with the heat transfer behavior of metal layer experiment (HELM) facility. The experiments are divided into two parts. Each part concerns 39 runs and 47 experimental conditions. Its corresponding results are obtained at middle Prandtl number (Pr = 7 for water) and the Nusselt number is found to be proportional to Ra 0.315 in the range 3.93 × 10 8 < Ra < 3.57 × 10 12 . Furthermore, the experiment

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Guide to data processing and revision: Part 2, Human error probability data entry and revision procedures

International Nuclear Information System (INIS)

Gilmore, W.E.; Gertman, D.I.; Gilbert, B.G.; Reece, W.J.

1988-11-01

The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability (HEP) and hardware component failure data (HCFD). The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. Users can perform data base searches to furnish HEP estimates and HCFD rates. In this manner, the NUCLARR system can be used to support a variety of risk assessment activities. This volume, Volume 3 of a 5-volume series, presents the procedures used to process HEP and HCFD for entry in NUCLARR and describes how to modify the existing NUCLARR taxonomy in order to add either equipment types or action verbs. Volume 3 also specifies the various roles of the administrative staff on assignment to the NUCLARR Clearinghouse who are tasked with maintaining the data base, dealing with user requests, and processing NUCLARR data. 5 refs., 34 figs., 3 tabs

Comparing imaging procedures: techniques and examples. Gastroenterology

International Nuclear Information System (INIS)

Malmud, L.S.

1982-01-01

The distinct advantages of nuclear medicine procedures, in comparison to radiography, contrast studies, computerized tomography and ultrasound, are emphasized. Scintigraphic methods offer quantitative data regarding function which competing imaging modalities are unable to provide, and make them the studies of choice in the evaluation of gastrointestinal physiology and functional abnormalities

Computerized tests to evaluate recovery of cognitive function after deep sedation with propofol and remifentanil for colonoscopy.

Science.gov (United States)

Borrat, Xavier; Ubre, Marta; Risco, Raquel; Gambús, Pedro L; Pedroso, Angela; Iglesias, Aina; Fernandez-Esparrach, Gloria; Ginés, Àngels; Balust, Jaume; Martínez-Palli, Graciela

2018-03-27

The use of sedation for diagnostic procedures including gastrointestinal endoscopy is rapidly growing. Recovery of cognitive function after sedation is important because it would be important for most patients to resume safe, normal life soon after the procedure. Computerized tests have shown being accurate descriptors of cognitive function. The purpose of the present study was to evaluate the time course of cognitive function recovery after sedation with propofol and remifentanil. A prospective observational double blind clinical study conducted in 34 young healthy adults undergoing elective outpatient colonoscopy under sedation with the combination of propofol and remifentanil using a target controlled infusion system. Cognitive function was measured using a validated battery of computerized cognitive tests (Cogstate™, Melbourne, Australia) at different predefined times: prior to starting sedation (Tbaseline), and then 10 min (T10), 40 min (T40) and 120 min (T120) after the end of colonoscopy. Tests included the assessment of psychomotor function, attention, visual memory and working memory. All colonoscopies were completed (median time: 26 min) without significant adverse events. Patients received a median total dose of propofol and remifentanil of 149 mg and 98 µg, respectively. Psychomotor function and attention declined at T10 but were back to baseline values at T40 for all patients. The magnitude of psychomotor task reduction was large (d = 0.81) however 100% of patients were recovered at T40. Memory related tasks were not affected 10 min after ending sedation. Cognitive impairment in attention and psychomotor function after propofol and remifentanil sedation was significant and large and could be easily detected by computerized cognitive tests. Even though, patients were fully recovered 40 min after ending the procedure. From a cognitive recovery point of view, larger studies should be undertaken to propose adequate criteria for discharge

Stratification of mammographic computerized analysis by BI-RADS categories

International Nuclear Information System (INIS)

Lederman, Richard; Leichter, Isaac; Buchbinder, Shalom; Novak, Boris; Bamberger, Philippe; Fields, Scott

2003-01-01

The Breast Imaging Reporting and Data System (BI-RADS) was implemented to standardize characterization of mammographic findings. The purpose of the present study was to evaluate in which BI-RADS categories the changes recommended by computerized mammographic analysis are most beneficial. Archival cases including, 170 masses (101 malignant, 69 benign) and 63 clusters of microcalcifications (MCs; 36 malignant, 27 benign), were evaluated retrospectively, using the BI-RADS categories, by several radiologists, blinded to the pathology results. A computerized system then automatically extracted from the digitized mammogram features characterizing mammographic lesions, which were used to classify the lesions. The results of the computerized classification scheme were compared, by receiver operating characteristics (ROC) analysis, to the conventional interpretation. In the ''low probability of malignancy group'' (excluding BI-RADS categories 4 and 5), computerized analysis improved the A z of the ROC curve significantly, from 0.57 to 0.89. In the ''high probability of malignancy group'' (mostly category 5) the computerized analysis yielded an ROC curve with an A z of 0.99. In the ''intermediate probability of malignancy group'' computerized analysis improved the A z significantly, from 0.66 for to 0.83. Pair-wise analysis showed that in the latter group the modifications resulting from computerized analysis were correct in 83% of cases. Computerized analysis has the ability to improve the performance of the radiologists exactly in the BI-RADS categories with the greatest difficulties in arriving at a correct diagnosis. It increased the performance significantly in the problematic group of ''intermediate probability of malignancy'' and pinpointed all the cases with missed cancers in the ''low probability'' group. (orig.)

Computerized industrial tomography

International Nuclear Information System (INIS)

Ashraf, M.M.

1999-01-01

Computerized Tomographic (CT) has been used for a number of applications in the field of medicine and industry. For the last couple of years, the technique has been applied for the material characterization and detection of defects and flaws inside the industrial components of nuclear, aerospace and missile industries. A CT scanner of first generation was developed at the institute. The scanner has been used to demonstrate couple of applications of CT in the field of non destructive testing of materials. The data acquired by placing the test objects at various angles and scanning the object through a source detector assembly has been processed on a Pentium computer for image reconstruction using a filtered back projection method. The technique has been developed which can be modified and improved to study various other applications in materials science and a modern computerized tomographic facility can be established. (author)

Uncertainty Evaluation of a Postulated LBLOCA for APR+ using KINS Realistic Evaluation Methodology and MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Hwang, Min Jeong; Marigomena, Ralph; Yoo, Tae Ho; Kim, Y. S.; Sim, S. K. [Environment and Energy Technology, Inc., Daejeon (Korea, Republic of); Bang, Young Seok [KINS, Daejeon (Korea, Republic of)

2014-05-15

As a part of the regulatory safety research, Korea Institute of Nuclear Safety(KINS) also developed a best estimate safety analysis regulatory audit code, MARS-KS, to realistically predict and better understand the physical phenomena of the design basis accidents. KINS improved uncertainty propagation methodology using MARS-KS and applied the improved uncertainty evaluation method for the Shinkori Units 3 and 4 LBLOC. This study is to evaluate the uncertainty propagation of a postulated LBLOCA and quantify the safety margin using KINS-REM and MARS-KS code for the APR+ (Advanced Pressurizer Reactor Plus) Standard Safety Analysis Report(SSAR) which is under regulatory review by the KINS for its design approval. KINS-REM LBLOCA realistic evaluation methodology was used for the regulatory assessment of the APR+ LBLOCA using MARS-KS to evaluate the uncertainty propagation of the uncertainty variables as well as to assess the safety margin during the limiting case of the APR+ double ended guillotine cold leg LBLOCA. Uncertainty evaluation for the APR+ LBLOCA shows that the reflood PCT with upper limit of 95% probability at 95% confidence level is 1363.2 K and is higher than the blowdown PCT95/95 of 1275.3 K. The result shows that the current evaluation of APR+ LBLOCA PCT is within the acceptance criteria of 1477 K ECCS.

Computerized tomography in orthopaedics and traumatology

International Nuclear Information System (INIS)

Boettger, E.; Heckl, R.; Rehabilitations-Krankenhaus Langensteinbach, Karlsbad

1981-01-01

Computerized tomography in traumatology is the selected method for the indications mentioned, so that angiographic investigations are only necessary in exceptional cases. Computerized tomography is also better than other methods when diagnozing soft part tumours, however, angiography is still indicated preoperatively for individual cases. CT is only good as additional help to conventional diagnostics with bone tumours. The differential diagnosis cysts-tumour is possible using contrast medium injections. The frequently large soft part share of tumours is recognizable with osteolytic tumours so that a better irradiation and operation planning can be effected. Diseases in the spinal canal can only be assessed with reservation using modern equipment. Lumbar dislocations of the disk can mostly not be sufficiently determined. Perivertebral abscesses can be certainly detected using computerized tomography. This is particularly so for abscesses prior to calcification. (orig.) [de

Major Achievements and Prospect of the ATLAS Integral Effect Tests

International Nuclear Information System (INIS)

Choi, K.; Kim, Y.; Song, C.; Baek, W.

2012-01-01

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R and D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

Radioactive airborne effluents and the environmental impact assessment of CAP1400 nuclear power plant under normal operation

Energy Technology Data Exchange (ETDEWEB)

Zhang, Qiong; Guo, RuiPing; Zhang, ChunMing; Chen, XiaoQiu; Wang, Bo, E-mail: wangbo@chinansc.cn

2014-12-15

Highlights: • Typical radionuclides dispersion from CAP1400 under normal operation was simulated. • Modified Gaussian model considered radioactive decay, dry and wet deposition and so on. • The radioactive impact pathways on the public through atmosphere were compared. • The maximum individual effective dose was lower than the public irradiation limit. - Abstract: China Advanced Passive nuclear power plant with installed capacity reaching to 1400 MW (CAP1400) is independently designed as the China's state-of-the-art third generation nuclear power brand based on AP1000 technology digestion and absorption. The concentration of typical radionuclides dispersed from CAP1400 under normal operation was calculated with modified Gaussian model, considering mixed layer height, dry deposition, wet deposition, radioactive decay and so on. The atmospheric dispersion factors, ground deposition rate, individual dose and public dose were also investigated to estimate the radioactive effects of CAP1400 under normal operation on surrounding environment and human beings. The radioactive impact pathways on the public through atmosphere, such as immersion irradiation in the smoke plume, internal irradiation from ingestion and inhalation and external irradiation from surface deposition were briefly introduced with focus on the comparison of the maximum individual effective dose to different group from atmospheric dispersion. All computation results show that the maximum individual irradiation dose happened to children with total effective irradiation dose of 4.52E−03 mSv/y, which was lower than the public irradiation limit of 0.25 mSv/y.

Radioactive airborne effluents and the environmental impact assessment of CAP1400 nuclear power plant under normal operation

International Nuclear Information System (INIS)

Zhang, Qiong; Guo, RuiPing; Zhang, ChunMing; Chen, XiaoQiu; Wang, Bo

2014-01-01

Highlights: • Typical radionuclides dispersion from CAP1400 under normal operation was simulated. • Modified Gaussian model considered radioactive decay, dry and wet deposition and so on. • The radioactive impact pathways on the public through atmosphere were compared. • The maximum individual effective dose was lower than the public irradiation limit. - Abstract: China Advanced Passive nuclear power plant with installed capacity reaching to 1400 MW (CAP1400) is independently designed as the China's state-of-the-art third generation nuclear power brand based on AP1000 technology digestion and absorption. The concentration of typical radionuclides dispersed from CAP1400 under normal operation was calculated with modified Gaussian model, considering mixed layer height, dry deposition, wet deposition, radioactive decay and so on. The atmospheric dispersion factors, ground deposition rate, individual dose and public dose were also investigated to estimate the radioactive effects of CAP1400 under normal operation on surrounding environment and human beings. The radioactive impact pathways on the public through atmosphere, such as immersion irradiation in the smoke plume, internal irradiation from ingestion and inhalation and external irradiation from surface deposition were briefly introduced with focus on the comparison of the maximum individual effective dose to different group from atmospheric dispersion. All computation results show that the maximum individual irradiation dose happened to children with total effective irradiation dose of 4.52E−03 mSv/y, which was lower than the public irradiation limit of 0.25 mSv/y

[Evaluation of the capacity of the APR-DRG classification system to predict hospital mortality].

Science.gov (United States)

De Marco, Maria Francesca; Lorenzoni, Luca; Addari, Piero; Nante, Nicola

2002-01-01

Inpatient mortality has increasingly been used as an hospital outcome measure. Comparing mortality rates across hospitals requires adjustment for patient risks before making inferences about quality of care based on patient outcomes. Therefore it is essential to dispose of well performing severity measures. The aim of this study is to evaluate the ability of the All Patient Refined DRG system to predict inpatient mortality for congestive heart failure, myocardial infarction, pneumonia and ischemic stroke. Administrative records were used in this analysis. We used two statistics methods to assess the ability of the APR-DRG to predict mortality: the area under the receiver operating characteristics curve (referred to as the c-statistic) and the Hosmer-Lemeshow test. The database for the study included 19,212 discharges for stroke, pneumonia, myocardial infarction and congestive heart failure from fifteen hospital participating in the Italian APR-DRG Project. A multivariate analysis was performed to predict mortality for each condition in study using age, sex and APR-DRG risk mortality subclass as independent variables. Inpatient mortality rate ranges from 9.7% (pneumonia) to 16.7% (stroke). Model discrimination, calculated using the c-statistic, was 0.91 for myocardial infarction, 0.68 for stroke, 0.78 for pneumonia and 0.71 for congestive heart failure. The model calibration assessed using the Hosmer-Leme-show test was quite good. The performance of the APR-DRG scheme when used on Italian hospital activity records is similar to that reported in literature and it seems to improve by adding age and sex to the model. The APR-DRG system does not completely capture the effects of these variables. In some cases, the better performance might be due to the inclusion of specific complications in the risk-of-mortality subclass assignment.

Stratification of mammographic computerized analysis by BI-RADS categories

Energy Technology Data Exchange (ETDEWEB)

Lederman, Richard [Department of Radiology, Hadassah University Hospital, Ein Kerem, Jerusalem (Israel); Leichter, Isaac [Department of Electro-Optics, Jerusalem College of Technology, P.O.B. 16031, Jerusalem (Israel); Buchbinder, Shalom [Department of Radiology of The Montefiore Medical Center, The University Hospital for the Albert Einstein College of Medicine, Bronx, New York (United States); Novak, Boris [Department of Applied Mathematics, Jerusalem College of Technology, P.O.B. 16031, Jerusalem 91160 (Israel); Bamberger, Philippe [Department of Electronics, Jerusalem College of Technology, POB 16031, Jerusalem (Israel); Fields, Scott [Department of Radiology, Hadassah University Hospital, Mt. Scopus, Jerusalem (Israel)

2003-02-01

The Breast Imaging Reporting and Data System (BI-RADS) was implemented to standardize characterization of mammographic findings. The purpose of the present study was to evaluate in which BI-RADS categories the changes recommended by computerized mammographic analysis are most beneficial. Archival cases including, 170 masses (101 malignant, 69 benign) and 63 clusters of microcalcifications (MCs; 36 malignant, 27 benign), were evaluated retrospectively, using the BI-RADS categories, by several radiologists, blinded to the pathology results. A computerized system then automatically extracted from the digitized mammogram features characterizing mammographic lesions, which were used to classify the lesions. The results of the computerized classification scheme were compared, by receiver operating characteristics (ROC) analysis, to the conventional interpretation. In the ''low probability of malignancy group'' (excluding BI-RADS categories 4 and 5), computerized analysis improved the A{sub z}of the ROC curve significantly, from 0.57 to 0.89. In the ''high probability of malignancy group'' (mostly category 5) the computerized analysis yielded an ROC curve with an A {sub z}of 0.99. In the ''intermediate probability of malignancy group'' computerized analysis improved the A {sub z}significantly, from 0.66 for to 0.83. Pair-wise analysis showed that in the latter group the modifications resulting from computerized analysis were correct in 83% of cases. Computerized analysis has the ability to improve the performance of the radiologists exactly in the BI-RADS categories with the greatest difficulties in arriving at a correct diagnosis. It increased the performance significantly in the problematic group of ''intermediate probability of malignancy'' and pinpointed all the cases with missed cancers in the ''low probability'' group. (orig.)

Computerized radiation treatment planning

International Nuclear Information System (INIS)

Laarse, R. van der.

1981-01-01

Following a general introduction, a chain consisting of three computer programs which has been developed for treatment planning of external beam radiotherapy without manual intervention is described. New score functions used for determination of optimal incidence directions are presented and the calculation of the position of the isocentre for each optimum combination of incidence directions is explained. A description of how a set of applicators, covering fields with dimensions of 4 to 20 cm, for the 6 to 20 MeV electron beams of a MEL SL75-20 linear accelerator was developed, is given. A computer program for three dimensional electron beam treatment planning is presented. A microprocessor based treatment planning system for the Selectron remote controlled afterloading system for intracavitary radiotherapy is described. The main differences in treatment planning procedures for external beam therapy with neutrons instead of photons is discussed. A microprocessor based densitometer for plotting isodensity lines in film dosimetry is described. A computer program for dose planning of brachytherapy is presented. Finally a general discussion about the different aspects of computerized treatment planning as presented in this thesis is given. (Auth.)

47 CFR 95.1121 - Specific requirements for wireless medical telemetry devices operating in the 1395-1400 and 1427...

Science.gov (United States)

2010-10-01

... wireless medical telemetry devices operating in the 1395-1400 and 1427-1432 MHz bands. Due to the critical... 47 Telecommunication 5 2010-10-01 2010-10-01 false Specific requirements for wireless medical telemetry devices operating in the 1395-1400 and 1427-1432 MHz bands. 95.1121 Section 95.1121...

The Alberta Pregnancy Outcomes and Nutrition (APrON) cohort study: rationale and methods.

Science.gov (United States)

Kaplan, Bonnie J; Giesbrecht, Gerald F; Leung, Brenda M Y; Field, Catherine J; Dewey, Deborah; Bell, Rhonda C; Manca, Donna P; O'Beirne, Maeve; Johnston, David W; Pop, Victor J; Singhal, Nalini; Gagnon, Lisa; Bernier, Francois P; Eliasziw, Misha; McCargar, Linda J; Kooistra, Libbe; Farmer, Anna; Cantell, Marja; Goonewardene, Laki; Casey, Linda M; Letourneau, Nicole; Martin, Jonathan W

2014-01-01

The Alberta Pregnancy Outcomes and Nutrition (APrON) study is an ongoing prospective cohort study that recruits pregnant women early in pregnancy and, as of 2012, is following up their infants to 3 years of age. It has currently enrolled approximately 5000 Canadians (2000 pregnant women, their offspring and many of their partners). The primary aims of the APrON study were to determine the relationships between maternal nutrient intake and status, before, during and after gestation, and (1) maternal mood; (2) birth and obstetric outcomes; and (3) infant neurodevelopment. We have collected comprehensive maternal nutrition, anthropometric, biological and mental health data at multiple points in the pregnancy and the post-partum period, as well as obstetrical, birth, health and neurodevelopmental outcomes of these pregnancies. The study continues to follow the infants through to 36 months of age. The current report describes the study design and methods, and findings of some pilot work. The APrON study is a significant resource with opportunities for collaboration. © 2012 John Wiley & Sons Ltd.

Veselības aprūpes iemaņas pusaudžiem bērnu namā

OpenAIRE

Zūpe, Daira

2008-01-01

Pētniecības darbā „Veselības aprūpes iemaņas pusaudžiem bērnu namā” autore Daira Zūpe izpēta un analizē vienu no bērnu namu aktuālajām problēmām veselības profilakses veicināšanā – veselības aprūpes iemaņu attīstīšanu. Pētījumā uzmanība pievērsta māsas lomai bērnu nama pusaudžu veselības aprūpes procesā, kur akcentēta veselības aprūpes pamatiemaņu nozīme veselības vecināšanā - fiziskās aktivitātes, ēšanas paradumi, kaitīgie ieradumi un higiēnas iemaņas. Pētniecības darba zinātniskā vadītā...

Aprūpes īpatnības bērniem ar Dauna sindromu stacionārā

OpenAIRE

Lūse, Ērika

2015-01-01

Bakalaura darba tēma – Aprūpes īpatnības bērniem ar Dauna sindromu stacionārā. Tēmas aktualitāte pievērst uzmanību bērna ar Dauna sindroma aprūpei, kā arī mazināt stereotipus par šiem bērniem. Darba mērķis ir izpētīt Dauna sindromu un māsas lomu šo bērnu aprūpē, kā arī aprūpes īpatnības. Pētniecības uzdevumi ir veikt zinātniskās literatūras analīzi, pētījuma atbilstošas māszinību teorijas atlasi, izstrādāt anketu, veikt pilot pētījumu, iegūto datu apkopošanu, analīze un secinājumu veikšana. I...

Integrated System Validation Usability Questionnaire: Computerized Procedures; Desarrollo del Cuestionario de Facilidad de Uso para la Validación de Sistemas Integrados: Procedimientos de Operacion

Energy Technology Data Exchange (ETDEWEB)

Garcés, Ma. I.; Torralba, B.

2015-07-01

The Research and Development (R&D) project on “Theoretical and Methodological Approaches to Integrated System Validation of Control Rooms, 2014-2015”, in which the research activities described in this report are framed, has two main objectives: to develop the items for an usability methodology conceived as a part of the measurement framework for performance-based control room evaluation that the OECD Halden Reactor Project will test in the experiments planned for 2015; and the statistical analysis of the data generated in the experimental activities of the Halden Man-Machine Laboratory (HAMMLAB) facility, with previous usability questionnaires, in 2010 and 2011. In this report, the procedure designed to meet the first goal of the project is described, in particular, the process followed to identify the items related to operating procedures, both computer and paper-based, one of the elements to be included in the usability questionnaire. Three phases are performed, in the first one, the approaches developed by the United States Nuclear Regulatory Commission, NRC, are reviewed, the models used by the nuclear industry and their technical support organizations, mainly, the Electric Power Research Institute, EPRI, are analyzed, and scientist advances are also explored. In the remaining stages, general and specific guidelines for computerized and paper-based procedures are compared and criteria for the preliminary selection of the items that should be incorporated into the usability questionnaire are defined. This proposal will be reviewed and adapted by the Halden Reactor Project to the design of the specific experiments performed in HAMLAB.

Research on accounting transition from computerization to informationization

Directory of Open Access Journals (Sweden)

Shu Chen

2017-11-01

Full Text Available The application for computer technology, digitalization technology and network technology in the accounting field has promoted the development of accounting informationization. Accounting informationization is a product integrated with traditional accounting theory and modern information technology, which is an inevitable trend of continuous development of modern accounting. This paper discusses the basic concepts and characteristics of accounting computerization and informationization based on the normative research method and literature data method, analyzes the feasibility of accounting transition from computerization to informationization, and finally puts forward the specific approaches and ultimate goals of accounting transition from computerization to informationization.

Patient radiation exposure in computerized tomography

Energy Technology Data Exchange (ETDEWEB)

Pavlov, V [Meditsinska Akademiya, Sofia (Bulgaria)

1980-01-01

Radiation exposure to patients undergoing axial computerized tomography as a tool of neurological X-ray diagnostics was studied. Doses thereby delivered were compared with those from routine head films at X-ray tube parameters 200 W, 70 kV, and 70 cm target-to-patient distance. Radiation exposures were analyzed with a view to improving shielding and procedural techniques. Comparisons were made using LiF TLD measurements with an Alderson phantom (standard for axial computer tomography). Skin and intracranial space doses were compared using two computers, Siretom I and Siretom 2000, for various positionings: frontal, fronto-lateral, temporal, temporo-occipital, and occipital. In addition, patient body doses with or without shielding and doses to subjects attending sick children or restless adults were examined. Achievable protection was estimated for lead shields of 0.5 mm lead equivalent. It was concluded that radiation doses delivered to neurologic patients undergoing axial computer tomography are smaller than those resulting from conventional X-ray examinations.

Cranial computerized tomography in children suffering from acute leukemia

International Nuclear Information System (INIS)

Metz, O.

1981-01-01

Cranial computerized (axial) tomography permits a more complete neurologic supervision of children with acute leukemia and a better knowledge of the frequency and varieties of cerebral complications in leukemia. Endocranial complications in acute leukemia are essentially infiltrative, hemorrhagic, infectious or iatrogenic. Cranial computerized tomography can demonstrate cerebral changes in meningeal leukemia, hemorrhages, calcifications, brain atrophy or leukencephalopathy. The preliminary results of cranial computerized tomography in childhood leukemia suggest that the iatrogenic main lesion of the brain due to combined radiation-chemotherapy is atrophy whereas that of the intrathecal cytostatic therapy is demyelination. Accurate diagnostics and control of possible cerebral complications in therapy of leukemia is essentially for appropriate therapeutic management. For that cranial computerized tomography is the best method to a effective supervision of the brain. (author)

Development of computerized supporting system for emergency response in nuclear power plant

International Nuclear Information System (INIS)

Lee, Jae Il

1992-02-01

In emergency situation of nuclear power plants, effective use of emergency operating procedures (EOPs) is a crucial part of the emergency response process. However, there are several problems in the emergency operating procedures because of the form of the written procedures. They are voluminous and complicate for effective references under high stress situation. Inevitably, it takes time that could be better spent employing measures to control and stabilize to select the correct procedures and apply the decision logic. In this study, a computerized supporting system has been developed to reduce the operator error possibility under emergency situations of nuclear power plant. Using on-line input parameters, the system can determine the status of the critical safety functions and can find appropriate procedures and necessary operator actions automatically. Moreover, the system can help the operator decision making in the core melt accident situation. By tracking the EOP in an on-line mode, most steps concerning checking or verifying plant state are processed automatically without operator participations. Therefore, the interactions between the system and the operator are simplified significantly and the possibility of human error is reduced

Apreçando Derivativos de Crédito no Brasil Credit Derivatives Pricing in Brazil

Directory of Open Access Journals (Sweden)

Jorge C. Kapotas

2004-12-01

Full Text Available Neste artigo apresentamos os principais modelos de apreçamento de instrumentos de crédito onde desenvolvem-se as equações de apreçamento de debêntures. Para tal, utilizam-se os modelos estruturais, de intensidade e, finalmente, de rating. A seguir, definimos o que são derivativos de crédito (CDS e derivamos fórmulas de apreçamento para estes instrumentos. Concluímos este artigo com uma aplicação ao mercado brasileiro. In this paper we present the main models used for pricing defaultable bonds and credit derivatives. The Merton structural model, the intensity framework and a Ratings based model are considered. We apply these techniques to the pricing of credit derivatives on Brazilian US$-indexed treasury bonds.

Computerized tabletop games as a form of a video game training for old-old.

Science.gov (United States)

Cujzek, Marina; Vranic, Andrea

2017-11-01

This research aimed at investigating the utility of a computerized version of a cognitively stimulating activity as a video game intervention for elderly. The study focused on the effect of a 6-week extensive practice intervention on aspects of cognitive functioning (vigilance, working memory (WM), inhibition, reasoning) of old-old participants (N = 29), randomly assigned to trained or active control group. The difference between groups was in the content of the extended video game practice - cognitively complex card game for trained and computerized version of a simple dice-game of chance for control participants. A pretest, posttest and a 4-month follow-up measurement was conducted. Results revealed improvements in both groups, except for improved reasoning found only in trained participants. These results suggest that: (1) improvements are dependent on the complexity of the program, (2) cognitively stimulating activity are a valid training procedure for old-old, (3) novelty of computer use is an important factor in determining training efficacy.

Computerized adaptive testing--ready for ambulatory monitoring?

DEFF Research Database (Denmark)

Rose, Matthias; Bjørner, Jakob; Fischer, Felix

2012-01-01

Computerized adaptive tests (CATs) have abundant theoretical advantages over established static instruments, which could improve ambulatory monitoring of patient-reported outcomes (PROs). However, an empirical demonstration of their practical benefits is warranted.......Computerized adaptive tests (CATs) have abundant theoretical advantages over established static instruments, which could improve ambulatory monitoring of patient-reported outcomes (PROs). However, an empirical demonstration of their practical benefits is warranted....

Écrire (d’après

Directory of Open Access Journals (Sweden)

Denis Saint-Amand

2009-03-01

Full Text Available De l’allusion à la citation, de l’emprunt criard au plagiat déguisé, les jeux d’influences et d’aller-retour avec tel ou tel modèle, grand ou mineur, ont toujours participé de l’élaboration du texte littéraire. Si l’intertextualité est sans aucun doute née en même temps que l’écriture (ou quelques minutes après et que, comme ont pu le montrer notamment Jacques-Philippe Saint-Gérand et Daniel Sangsue, les facettes multiples qu’elle déploie ont toujours suscité l’attention des hommes de lettre...

Computerized systems for high level information processing and decision making in the field of PSA

International Nuclear Information System (INIS)

Kafka, P.; Kunitz, H.

1990-01-01

A comprehensive review of Probabilistic Safety Assessment (PSA) related program packages is made. Three fields in methodological succession are covered: plant modeling, data quantification procedures and decision-making support. Packages for fault tree construction and minimal cut sets evaluation are referred to and the performances of three of them: RALLY, ORCHARD and SALP-PC, are discussed and compared. Notions on the raw data sources are presented and the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is given as an example for data base management system generating PSA data. Aggregated risk models for support in safety assessment, plant operation and accident management (SARA, ESSM, PRISIM) are cited. Examples of systems supporting 'living PSA' (SAIS, SUPER-NET, LESSEPS 1300, NUPRA, SPSA) are given. The concluding remarks outline the state-of-the-art developments of computerized systems for reliability analyses. 1 fig., 1 tab., 51 refs. (R.Ts)

New tools for NTD vaccines: A case study of quality control assays for product development of the human hookworm vaccine Na-APR-1M74.

Science.gov (United States)

Pearson, Mark S; Jariwala, Amar R; Abbenante, Giovanni; Plieskatt, Jordan; Wilson, David; Bottazzi, Maria Elena; Hotez, Peter J; Keegan, Brian; Bethony, Jeffrey M; Loukas, Alex

2015-01-01

Na-APR-1(M74) is an aspartic protease that is rendered enzymatically inactive by site-directed mutagenesis and is a candidate antigen component in the Human Hookworm Vaccine. The mutant protease exerts vaccine efficacy by inducing antibodies that neutralize the enzymatic activity of wild type enzyme (Na-APR-1wt) in the gut of the hookworm, thereby depriving the worm of its ability to digest its blood meal. Previously, canines immunized with Na-APR-1(M74) and challenged with Ancylostoma caninum were partially protected against hookworm challenge infection, especially from the loss in hemoglobin observed in control canines and canine immunoglobulin (Ig) G raised against Na-APR-1 was shown to inhibit the enzymatic activity of Na-APR-1 wt in vitro, thereby providing proof of concept of Na-APR-1(M74) as a vaccine antigen. The mutated version, Na-APR-1(M74), was then expressed at the cGMP level using a Nicotiana benthamiana expression system (Fraunhofer, CMB, Delaware, MD), formulated with Alhydrogel®, and used to immunize mice in a dose-ranging study to explore the enzyme-neutralizing capacity of the resulting anti- Na-APR-1(M74) IgG. As little as 0.99 μg of recombinant Na-APR-1(M74) could induce anti Na-APR-1(M74) IgG in mice that were capable of inhibiting Na-APR-1w t-mediated digestion of a peptide substrate by 89%. In the absence of enzymatic activity of Na-APR-1(M74) as a surrogate marker of protein functionality, we developed an assay based on the binding of a quenched fluorescence-labeled inhibitor of aspartic proteases, BODIPY-FL pepstatin A (BDP). Binding of BDP in the active site of Na-APR-1 wt was demonstrated by inhibition of enzymatic activity, and competitive binding with unlabelled pepstatin A. BDP also bound to Na-APR-1(M74) which was assessed by fluorescence polarization, but with an ∼ 50-fold reduction in the dissociation constant. Taken together, these assays comprise a "toolbox" that could be useful for the analyses of Na-APR-1(M74) as it

Procedures simulation in Nuclear Power Plants (SIMPROC); Codigo para la simulacion de procedimientos humanos en centrales nucleares: SIMPROC

Energy Technology Data Exchange (ETDEWEB)

Izquierdo, J. M.; Hortal, J.; Sanchez, M.; Melendez, E.; Queral, C.; Ibanez, L.; Gil, J.; Fernandez, I.; Murcia, S.; Gomez, J.

2010-07-01

SIMPROC, a procedure simulator support in rules, is an Indizen product whose procedure computerization is based in XML. It's a compliment for the traditional methods of procedures evaluation and it's connected to SCAIS (Code System for an Integrate Security Analysis).

26 CFR 1.1400L(b)-1 - Additional first year depreciation deduction for qualified New York Liberty Zone property.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 11 2010-04-01 2010-04-01 true Additional first year depreciation deduction for... to Individuals' Title 11 Cases § 1.1400L(b)-1 Additional first year depreciation deduction for...-percent additional first year depreciation deduction allowable under section 1400L(b) for qualified New...

Computerized data base of the fundamental constants of nature

International Nuclear Information System (INIS)

Henry, E.A.; Hampel, V.E.

1975-01-01

Fifty-seven fundamental constants of nature were computerized from the up-to-date evaluations of E. R. Cohen and B. N. Taylor. The constants are annotated with regard to symbol, value, uncertainty, and scaling factor. This computerization is part of the scientific data base project of the Information Research Group at Lawrence Livermore Laboratory. The MASTER CONTROL data base management system is used. The computerized fundamental constants can be requested from the ERDA Computer Program Exchange and Information Center of the Argonne National Laboratory or from the National Technical Information Service of the U. S. Department of Commerce. This is the first of a series of releases on preparation of computerized scientific and technological data banks. The next release is a data bank of conversion factors for different units of measurements. 3 figures

Evaluation of ATLAS 100% DVI Line Break Using TRACE Code

International Nuclear Information System (INIS)

Huh, Byung Gil; Bang, Young Seok; Cheong, Ae Ju; Woo, Sweng Woong

2011-01-01

ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is an integral effect test facility in KAERI. It had installed completely to simulate the accident for the OPR1000 and the APR1400 in 2005. After then, several tests for LBLOCA, DVI line break have been performed successfully to resolve the safety issues of the APR1400. Especially, a DVI line break is considered as another spectrum among the SBLOCAs in APR1400 because the DVI line is directly connected to the reactor vessel and the thermal hydraulic behaviors are expected to be different from those for the cold leg injection. However, there are not enough experimental data for the DVI line break. Therefore, integral effect data for the DVI line break of ATLAS is very useful and available for an improvement and validation of safety codes. For the DVI line break in ATLAS, several analyses using MARS and RELAP codes were performed in the ATLAS DSP (Domestic Standard Problem) meetings. However, TRACE code has still not used to simulate a DVI line break. TRACE code has developed as the unified code for the reactor thermal hydraulic analyses in USNRC. In this study, the 100% DVI line break in ATLAS was evaluated by TRACE code. The objectives of this study are to identify the prediction capability of TRACE code for the major thermal hydraulic phenomena of a DVI line break in ATLAS

Five Tubes Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-CL-02 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Compared to the case of a single U-tube rupture test, opening frequency of the MSSVs in the intact steam generator (SG-2) was highly reduced after 500 seconds in the present SGTR-CL-02 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

Five Tubes Rupture at Hot Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-HL-05 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the SPACE code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. On the contrary to the case of a single U-tube rupture test, the MSSV of the intact steam generator was not opened any more after 1500 seconds in the present SGTR-HL-05 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

Computerized videodefecography versus defecography: do we need radiographs?

Directory of Open Access Journals (Sweden)

Carlos Walter Sobrado

Full Text Available CONTEXT AND OBJECTIVE: Defecography has been recognized as a valuable method for evaluating patients with evacuation disorders. It consists of the use of static radiography and fluoroscopy to record different situations within anorectal dynamics. Conventionally, rectal parameters are measured using radiograms. It is rare for fluoroscopy alone to be used. Computer software has been developed with the specific aim of calculating these measurements from digitized videotaped images obtained during fluoroscopy, without the need for radiographic film, thereby developing a computerized videodefecography method. The objective was thus to compare measurements obtained via computerized videodefecography with conventional measurements and to discuss the advantages of the new method. DESIGN AND SETTING: Prospective study at the radiology service of Hospital das Clínicas, Universidade de São Paulo. METHOD: Ten consecutive normal subjects underwent videodefecography. The anorectal angle, anorectal junction, puborectalis muscle length, anal canal length and degree of anal relaxation were obtained via the conventional method (using radiography film and via computerized videodefecography using the ANGDIST software. Measurement and analysis of these parameters was performed by two independent physicians. RESULTS: Statistical analysis confirmed that the measurements obtained through direct radiography film assessment and using digital image analysis (computerized videodefecography were equivalent. CONCLUSIONS: Computerized videodefecography is equivalent to the traditional defecography examination. It has the advantage of offering reduced radiation exposure through saving on the use of radiography.

Country Report for Republic of Korea

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Woon [Shinil General Hospital, 47-7, Suyou5dong, Kangbuk-Gu, 142-070 Suyou 5 Dong (Korea, Republic of)

2008-07-01

Nuclear Power in Korea: Electric generation from 20 reactors: 130,715 million kWh/y; Nuclear generation of total electricity: 40 %. Reactor types: - 4 PHWR (Candu), 16 PWR (6 USA-Westinghouse, 2 French-Framatome, 8 Korean-OPR1000). Nuclear future: - Under Construction: 4 PWR (OPR1000), 2 PWR (APR1400, ShinKori no 3 and 4). - Planned: 2 PWR (APR1400, ShinWooljin no 3 and 4). - R and D investment plan in 2008 managed by MEST (ex- MOST) due to new governmental unification: - next generation reactor: 2 units of APR1400 construction and development of APR1400+, - NTP cooperative research: Gen-IV with SFR, Pyro-process, - Nuclear application: SMART, KSTAR for ITER construction, - Waste disposal management: LILW, SF, Decommission, - RFT, Safety enforcement, radiation protection, public communication, medical use, etc. Nuclear Power in Korea: Status of Radioactive Waste Disposal Site: - Site: Kyeongju, southeast part of Korea, - Size: 2,098,419 m{sup 2}, - Waste type: Low/Intermediate Level Waste (LILW) and RI waste, - Disposal type: Cavern, - Capacity: 1. stage 100,000 drums (total 800,000 drums), - Under Licensing Process for Construction WIN-Korea Activity in 2007: 250 members and 11 organization sponsors. 1. 2007 Symposium on Nuclear with Women: - Topic: 'Rendezvous of Nuclear and Renewable Energy', October 11, 2007 at Conference Hall of National Congress. Audience: 400 from Women NGO and the Assembly. - Panel: six specialists from energy, journalism, and women NGO's discussed about energy secure, saving, development of environment friendly energy, protection from global warming. - Results: public communication effect from mass media. 2. 2007 Symposium Conferences: General Meeting/workshop, Nov. 6, 2007, sponsored by KOFWST (Korea Federation of Women's Science and Technology Associations)/Seoul Education Culture Centre. participants: 100 members. Lectures: 'Women's Self Leadership', 'Survey Results of Public understanding on

Highly resolving computerized tomography

International Nuclear Information System (INIS)

Kurtz, B.; Petersen, D.; Walter, E.

1984-01-01

With the development of highly-resolving devices for computerized tomography, CT diagnosis of the lumbar vertebral column has gained increasing importance. As an ambulatory, non-invasive method it has proved in comparative studies to be at least equivalent to myelography in the detection of dislocations of inter-vertebral disks (4,6,7,15). Because with modern devices not alone the bones, but especially the spinal soft part structures are clearly and precisely presented with a resolution of distinctly below 1 mm, a further improvement of the results is expected as experience will increase. The authors report on the diagnosis of the lumbar vertebral column with the aid of a modern device for computerized tomography and wish to draw particular attention to the possibility of doing this investigation as a routine, and to the diagnostic value of secondary reconstructions. (BWU) [de

Highly resolving computerized tomography

Energy Technology Data Exchange (ETDEWEB)

Kurtz, B.; Petersen, D.; Walter, E.

1984-01-01

With the development of highly-resolving devices for computerized tomography, CT diagnosis of the lumbar vertebral column has gained increasing importance. As an ambulatory, non-invasive method it has proved in comparative studies to be at least equivalent to myelography in the detection of dislocations of inter-vertebral disks (4,6,7,15). Because with modern devices not alone the bones, but especially the spinal soft part structures are clearly and precisely presented with a resolution of distinctly below 1 mm, a further improvement of the results is expected as experience will increase. The authors report on the diagnosis of the lumbar vertebral column with the aid of a modern device for computerized tomography and wish to draw particular attention to the possibility of doing this investigation as a routine, and to the diagnostic value of secondary reconstructions.

897.pdf | apr102004 | currsci | Indian Academy of Sciences

Indian Academy of Sciences (India)

Home; currsci; apr102004; 897.pdf. 404! error. The page your are looking for can not be found! Please check the link or use the navigation bar at the top. YouTube; Twitter; Facebook; Blog. Academy News. IAS Logo. 29th Mid-year meeting. Posted on 19 January 2018. The 29th Mid-year meeting of the Academy will be held ...

Computerized tomography in evaluation of the pulmonary hilum

International Nuclear Information System (INIS)

Secaf, M.; Ferreira, J.L.N.; Secaf, E.

1987-01-01

The use of computerized tomography as a method for evaluating the pulmonary hilum and its application in the diagnosis of endobronchial lesions, hilar masss, contiguous involvement of the mediastinum by hilar mass, and vascular hilar lesions are discussed. A comparative evaluation between conventional tomograms and computerized tomography is presented. (M.A.C.) [pt

Seismic response analysis of a piping system subjected to multiple support excitations in a base isolated NPP building

International Nuclear Information System (INIS)

Surh, Han-Bum; Ryu, Tae-Young; Park, Jin-Sung; Ahn, Eun-Woo; Choi, Chul-Sun; Koo, Ja Choon; Choi, Jae-Boong; Kim, Moon Ki

2015-01-01

Highlights: • Piping system in the APR 1400 NPP with a base isolation design is studied. • Seismic response of piping system in base isolated building are investigated. • Stress classification method is examined for piping subjected to seismic loading. • Primary stress of piping is reduced due to base isolation design. • Substantial secondary stress is observed in the main steam piping. - Abstract: In this study, the stress response of the piping system in the advanced power reactor 1400 (APR 1400) with a base isolation design subjected to seismic loading is addressed. The piping system located between the auxiliary building with base isolation and the turbine building with a fixed base is considered since it can be subjected to substantial relative support movement during seismic events. First, the support responses with respect to the base characteristic are investigated to perform seismic analysis for multiple support excitations. Finite element analyses are performed to predict the piping stress response through various analysis methods such as the response spectrum, seismic support movement and time history method. To separately evaluate the inertial effect and support movement effect on the piping stress, the stress is decomposed into a primary and secondary stress using the proposed method. Finally, influences of the base isolation design on the piping system in the APR 1400 are addressed. The primary stress based on the inertial loading is effectively reduced in a base isolation design, whereas a considerable amount of secondary stress is generated in the piping system connecting a base isolated building with a fixed base building. It is also confirmed that both the response spectrum analysis and seismic support movement analysis provide more conservative estimations of the piping stress compared to the time history analysis

Implementasi Automatic Packet Reporting System (APRS Untuk Paket Data Pemantauan dan PengukuranUntuk Paket Data Pemantauan dan Pengukuran

Directory of Open Access Journals (Sweden)

Arief Goeritno

2016-03-01

Full Text Available Telah dilakukan implementasi Automatic Packet Reporting System (APRS untuk paket data pemantauan dan pengukuran melalui tujuan penelitian, berupa: a penyetelan program aplikasi pada jaringan APRS dan b pengukuran terhadap penerimaan data berdasarkan kinerja sensor-sensor. Penyetelan program aplikasi APRS merupakan konfigurasi perangkat lunak untuk APRS yang akan digunakan pada stasiun penerimaan data APRS dengan program aplikasi yang biasa digunakan, yaitu hyperterminal dan UI-View 32. Pengukuran penerimaan data berdasarkan kinerja sensor yang dilakukan melalui proses perekaman pada stasiun penerimaan data APRS. Kinerja sensor-sensor akan diamati pada stasiun pengiriman dan data hasil pengamatan akan dapat diterima pada stasiun penerimaan secara real time. Program aplikasi berbasis hyperterminal dan UI View 32 telah berhasil melakukan proses handshaking antara Terminal Node Controller (TNC dan komputer, sehingga data telemetri dari stasiun pengiriman paket data dapat diterima di stasiun penerimaan. Data telemetri dapat diamati pada stasiun penerimaan dan dapat diperoleh secara real time dengan format: YB0LRB-11>BEACON,WIDE2-1 [05/18/2014 04:03:49]: : T#010,008,093,004,122,075. Notifikasi YB0LRB-11 merupakan stasiun pengiriman paket data telemetri, kemudian data tersebut akan diterima pada stasiun penerima YD1PRY dengan format: YD1PRY-2>APLPN,ARISS [05/18/2014 04:03:07]: : !06.30.37S/106.48.26E#. Notifikasi tersebut merupakan pengiriman informasi data posisi oleh stasiun YD1PRY untuk inisialisasi pada jaringan APRS. Stasiun YB0LRB-11 ketika mengirim paket data telemetri dengan format: YB0LRB-11>BEACON,WIDE2-1 [05/18/2014 04:00:49]: : T#001,004,035,005,122,075. Paket data dari stasiun YB0LRB yang dipancar ulang atau digipeater dengan format: YB0LRB-11>BEACON,YD1PRY-2,WIDE2* [05/18/2014 04:00:50]: : T#001,004,035,005,122,075. Sensor pengukuran berkinerja relatif stabil, walaupun terdapat nilai simpangan pengukuran sebesar 1 cm atau

Celebral computerized tomography

International Nuclear Information System (INIS)

Lofteroed, B.; Sortland, O.

1985-01-01

Indications for cerebral computerized tomography (CT) and the diagnostic results from this examination are evaluated in 127 children. Pathological changes were found in 31 children, mostly based on such indications as increasing head size, suspicion of brain tumor, cerebral paresis, delayed psychomotor development and epileptic seizures. A list of indications for CT in children is given

Survey of methods for improving operator acceptance of computerized aids

International Nuclear Information System (INIS)

Frey, P.R.; Kisner, R.A.

1982-04-01

The success of current attempts to improve the operational performance and safety of nuclear power plants by installing computerized operational aids in the control rooms is dependent, in part, on the operator's attitude toward the aid. Utility experience with process computer systems indicates that problems may already exist with operator acceptance of computerized aids. The growth of the role that computers have in nuclear power plants makes user acceptance of computer technology an important issue for the nuclear industry. The purpose of this report is to draw from the literature factors related to user acceptance of computerized equipment that may also be applicable to the acceptance of computerized aids used in the nuclear power plant control room

Survey of methods for improving operator acceptance of computerized aids

Energy Technology Data Exchange (ETDEWEB)

Frey, P. R.; Kisner, R. A.

1982-04-01

The success of current attempts to improve the operational performance and safety of nuclear power plants by installing computerized operational aids in the control rooms is dependent, in part, on the operator's attitude toward the aid. Utility experience with process computer systems indicates that problems may already exist with operator acceptance of computerized aids. The growth of the role that computers have in nuclear power plants makes user acceptance of computer technology an important issue for the nuclear industry. The purpose of this report is to draw from the literature factors related to user acceptance of computerized equipment that may also be applicable to the acceptance of computerized aids used in the nuclear power plant control room.

Computerized accounting methods. Final report

International Nuclear Information System (INIS)

1994-01-01

This report summarizes the results of the research performed under the Task Order on computerized accounting methods in a period from 03 August to 31 December 1994. Computerized nuclear material accounting methods are analyzed and evaluated. Selected methods are implemented in a hardware-software complex developed as a prototype of the local network-based CONMIT system. This complex has been put into trial operation for test and evaluation of the selected methods at two selected ''Kurchatov Institute'' Russian Research Center (''KI'' RRC) nuclear facilities. Trial operation is carried out since the beginning of Initial Physical Inventory Taking in these facilities that was performed in November 1994. Operation of CONMIT prototype system was demonstrated in the middle of December 1994. Results of evaluation of CONMIT prototype system features and functioning under real operating conditions are considered. Conclusions are formulated on the ways of further development of computerized nuclear material accounting methods. The most important conclusion is a need to strengthen computer and information security features supported by the operating environment. Security provisions as well as other LANL Client/Server System approaches being developed by Los Alamos National Laboratory are recommended for selection of software and hardware components to be integrated into production version of CONMIT system for KI RRC

75 FR 44847 - Proposed Collection; Comment Request for Forms 943, 943-PR, 943-A, and 943A-PR

Science.gov (United States)

2010-07-29

... 943, 943-PR, 943- A, and 943A-PR AGENCY: Internal Revenue Service (IRS), Treasury. ACTION: Notice and... comments concerning Forms 943, Employer's Annual Tax Return for Agricultural Employees, 943-PR, Planilla..., Agricultural Employer's Record of Federal Tax Liability, and 943A-PR, Registro De La Obligacion Contributiva...

The Green Bank Third (GB3) survey of extragalactic radio sources at 1400 MHz

International Nuclear Information System (INIS)

Rys, S.; Machalski, J.

1987-01-01

The NRAO 91-m radio telescope in Green Bank, West Virginia was used to make a 1400-MHz sky survey covering an area of 0.0988 sr at declinations 70 deg ≤ δ 1950 < 76.8 deg with 10.1 x 10.5 arcmin resolution. This survey ends the series of smaller than 1-sr surveys made at 1400 MHz with that telescope and four-feed radiometer. A catalogue of 502 radiosources is presented, statistically complete to 112 mJy, which is about five times the rms noise and extragalactic confusion. The observations and data reduction are briefly summarized; the position and flux density errors are discussed. 13 refs., 2 tabs. (author)

Procedural-support music therapy in the healthcare setting: a cost-effectiveness analysis.

Science.gov (United States)

DeLoach Walworth, Darcy

2005-08-01

This comparative analysis examined the cost-effectiveness of music therapy as a procedural support in the pediatric healthcare setting. Many healthcare organizations are actively attempting to reduce the amount of sedation for pediatric patients undergoing various procedures. Patients receiving music therapy-assisted computerized tomography scans ( n = 57), echocardiograms ( n = 92), and other procedures ( n = 17) were included in the analysis. Results of music therapy-assisted procedures indicate successful elimination of patient sedation, reduction in procedural times, and decrease in the number of staff members present for procedures. Implications for nurses and music therapists in the healthcare setting are discussed.

78 FR 47051 - Proposed Collection; Comment Request for Forms 943, 943-PR, 943-A, and 943A-PR

Science.gov (United States)

2013-08-02

... 943, 943-PR, 943- A, and 943A-PR AGENCY: Internal Revenue Service (IRS), Treasury. ACTION: Notice and..., 943-PR, Planilla Para La Declarcion Anual De La Contribucion Federal Del Patrono De Empleados Agricolas, 943-A, Agricultural Employer's Record of Federal Tax Liability, and 943A-PR, Registro De La...

76 FR 11944 - Standard Instrument Approach Procedures, and Takeoff Minimums and Obstacle Departure Procedures...

Science.gov (United States)

2011-03-04

......... 1/3821 2/7/11 RNAV (GPS) RWY 19, Orig 7-Apr-11 SC Columbia Jim Hamilton L.B. 1/4017 2/8/11 GPS RWY.../11 RNAV (GPS) RWY 3, Orig 7-Apr-11 IL Chicago Chicago-O'Hare Intl. 0/7506 2/7/11 ILS OR LOC RWY 9L..... 0/8316 2/7/11 RNAV (GPS) RWY 15, Orig 7-Apr-11 NE Nebraska City....... Nebraska City Muni.. 0/8318 2...

MAPS evaluation report and procedures governing interviews and performance appraisals

CERN Multimedia

HR Department

2006-01-01

Following various improvements made to the MAPS report and to the procedures governing interviews and performance appraisals (announced in the CERN Weekly Bulletin 48-49/2005), three information sessions have been organized for all staff members: 24 January 10:00: AB Auditorium P (864-1-D02), 26 January 14:00: Main Amphitheatre, 31 January 10:00: AB Auditorium P (864-1-D02). Human Resources Department Tel. 73566

GPM GROUND VALIDATION AIRBORNE SECOND GENERATION PRECIPITATION RADAR (APR-2) GCPEX V1

Data.gov (United States)

National Aeronautics and Space Administration — The GPM Ground Validation Airborne Second Generation Precipitation Radar (APR-2) GCPEx dataset was collected during the GPM Cold-season Precipitation Experiment...

1001.pdf | 25apr2010 | currsci | Indian Academy of Sciences

Indian Academy of Sciences (India)

2010-04-25

Apr 25, 2010 ... error. The page your are looking for can not be found! Please check the link or use the navigation bar at the top. YouTube · Twitter · Facebook · Blog. Academy News. IAS Logo. Summer Research Fellowship Programme 2018. Dates Extended To 7 December 2017. Register here · Focus Area Science ...

45 CFR 307.25 - Review and certification of computerized support enforcement systems.

Science.gov (United States)

2010-10-01

..., DEPARTMENT OF HEALTH AND HUMAN SERVICES COMPUTERIZED SUPPORT ENFORCEMENT SYSTEMS § 307.25 Review and certification of computerized support enforcement systems. The Office will review, assess and inspect the... 45 Public Welfare 2 2010-10-01 2010-10-01 false Review and certification of computerized support...

Computerized 50 liter volume calibration system

International Nuclear Information System (INIS)

Proffitt, T.H.

1990-01-01

A system has been designed for the Savannah River Site that will be used to calibrate product shipping containers. For accountability purposes, it is necessary that these containers be calibrated to a very high precision. The Computerized 50 Liter Volume Calibration System (CVCS), which is based on the Ideal Gas Law (IGL), will use reference volumes with precision of no less ±0.03%, and helium to calibrate the containers to have a total error of no greater than ±0.10%. A statistical interpretation of the system has given a theoretical total calculated error of ±0.08%. Tests with the system will be performed once fabrication is complete to experimentally verify the calculated error. Since the total error was calculated using the worst case scenario, the actual error should be significantly less than the calculated value. The computer controlled, totally automated system is traceable to the National Institute of Standards and Technology. The design, calibration procedure, and statistical interpretation of the system will be discussed. 1 ref

Preliminary Evaluations of CSPACE for a Station Blackout Transient in APR1400

Energy Technology Data Exchange (ETDEWEB)

Lee, T. B.; Lee, D. K.; Lee, H. S.; Lee, G. W.; Choi, T. S. [KEPCO, Daejeon (Korea, Republic of); Park, R. J.; Kim, D. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

This paper discusses the preliminary results of the simulated station blackout (SBO) transients using the CSPACE code and presents the information pertinent to the related safety issues. CSPACE is a merged program of a master processer of Safety and Performance Analysis Code (SPACE) for nuclear power plants and a child processer of Core Meltdown Progression Accident Simulation Software (COMPASS) generated as a dynamic-link library (DLL) codes. It has been developed to predict the best-estimate transient in the pressurized water reactor (PWR) for severe accidents. SPACE and COMPASS codes take charge of the thermal-hydraulic response of PWRs and the analysis of the severe accident progression in a vessel, respectively. The initial phase is estimated starting from time zero when the loss of off-site and on-site powers occurs simultaneously. Shortly after the RCS pressure initially falls and rises slightly due to the effects of the reactor and turbine trips, the RCS pressure declines in response to the cooling provided by heat removed to the SGs. During the period of the primary heat-up and boil-off, the RCS pressure increase is limited by two cycles of the POSRV. The RCS fluid mass is lost through the pressurizer POSRV and then the core uncovers and superheated steam flows out from the RV into the coolant loops starting at 5513.0 seconds.

76 FR 16686 - Standard Instrument Approach Procedures, and Takeoff Minimums and Obstacle Departure Procedures...

Science.gov (United States)

2011-03-25

... Butler Butler County/K W 1/4580 2/3/11 ILS or LOC RWY 8, Amdt 7B Scholter Field. 7-Apr-11 FL Orlando Orlando Sanford 1/4606 2/3/11 ILS or LOC RWY 9L, Amdt 3 Intl. 7-Apr-11 FL Orlando Orlando Sanford 1/4608 2/3/11 NDB B, Orig Intl. 7-Apr-11 FL Orlando Orlando Sanford 1/4609 2/3/11 ILS or LOC RWY 27R, Amdt 1...

Plastic deformation of cubic zirconia single crystals at 1400 C

International Nuclear Information System (INIS)

Baufeld, B.; Baither, D.; Bartsch, M.; Messerschmidt, U.

1998-01-01

Cubic zirconia single crystals stabilized with 11 mol% yttria were deformed in air at 1400 C and around 1200 C at different strain rates along [1 anti 12] and [100] compression directions. The strain rate sensitivity of the flow stress was determined by strain rate cycling and stress relaxation tests. The microstructure of the deformed specimens was investigated by transmission high-voltage electron microscopy, including contrast extinction analysis for determining the Burgers vectors as well as stereo pairs and wide-angle tilting experiments to find the active slip planes. At deformation along [1 anti 12], the primary and secondary slip planes are of {100} type. Previous experiments had shown that the dislocations move easily on these planes in an athermal way. During deformation along [100], mainly dislocations on {100} planes are activated, which move in a viscous way by the aid of thermal activation. The discussion of the different deformation behaviours during deformation along [1 anti 12] and [100] is based on the different dynamic properties of dislocations and the fact that recovery is an essential feature of the deformation of cubic zirconia at 1400 C. The results on the shape of the deformation curve and the strain rate sensitivity of the flow stress are partly at variance with those of previous authors. (orig.)

Computerized Virtual Reality Simulation in Preclinical Dentistry: Can a Computerized Simulator Replace the Conventional Phantom Heads and Human Instruction?

Science.gov (United States)

Plessas, Anastasios

2017-10-01

In preclinical dental education, the acquisition of clinical, technical skills, and the transfer of these skills to the clinic are paramount. Phantom heads provide an efficient way to teach preclinical students dental procedures safely while increasing their dexterity skills considerably. Modern computerized phantom head training units incorporate features of virtual reality technology and the ability to offer concurrent augmented feedback. The aims of this review were to examine and evaluate the dental literature for evidence supporting their use and to discuss the role of augmented feedback versus the facilitator's instruction. Adjunctive training in these units seems to enhance student's learning and skill acquisition and reduce the required faculty supervision time. However, the virtual augmented feedback cannot be used as the sole method of feedback, and the facilitator's input is still critical. Well-powered longitudinal randomized trials exploring the impact of these units on student's clinical performance and issues of cost-effectiveness are warranted.

Automatic Control of Reactor Temperature and Power Distribution for a Daily Load following Operation

Energy Technology Data Exchange (ETDEWEB)

Yu, Keuk Jong; Kim, Han Gon [Korea Hydro and Nuclear Power Institute, Daejeon (Korea, Republic of)

2010-10-15

An automatic control method of reactor power and power distribution was developed for a daily load following operation of APR1400. This method used a model predictive control (MPC) methodology having second-order plant data. And it utilized a reactor power ratio and axial shape index as control variables. However, the reactor regulating system of APR1400 is operated by the difference between the average temperature of the reactor core and the reference temperature, which is proportional to the turbine load. Thus, this paper reports on the model predictive control methodology using fourth-order plant data and a reactor temperature instead of the reactor power shape. The purpose of this study is to develop a revised automatic controller and analyze the behavior of the nuclear reactor temperature (Tavg) and the axial shape index (ASI) using the MPC method during a daily load following operation

A Computerized QC Analysis of TLD Glow Curves for Personal Dosimetry Measurements Using Tag QC Program

International Nuclear Information System (INIS)

Primo, S.; Datz, H.; Dar, A.

2014-01-01

The External Dosimetry Lab (EDL) at the Radiation Safety Division at Soreq Nuclear Research Center (SNRC) is ISO 17025 certified and provides its services to approximately 13,000 users throughout the country from various sectors such as medical, industrial and academic. About 95% of the users are monitored monthly for X-rays, radiation using Thermoluminescence Dosimeter (TLD) cards that contain three LiF:Mg,Ti elements and the other users, who work also with thermal neutrons, use TLD cards that contain four LiF:Mg,Ti elements. All TLD cards are measured with the Thermo 8800pc reader. Suspicious TLD glow curve (GC) can cause wrong dose estimation so the EDL makes great efforts to ensure that each GC undergoes a careful QC procedure. The current QC procedure is performed manually and through a few steps using different softwares and databases in a long and complicated procedure: EDL staff needs to export all the results/GCs to be checked to an Excel file, followed by finding the suspicious GCs, which is done in a different program (WinREMS), According to the GC shapes (Figure 1 illustrates suitable and suspicious GC shapes) and the ratio between the elements result values, the inspecting technician corrects the data. The motivation for developing the new program is the complicated and time consuming process of our the manual procedure to the large amount of TLDs each month (13,000), similarly to other Dosimetry services that use computerized QC GC analysis. it is important to note that only ~25% of the results are above the EDL recorded level (0.10 mSv) and need to be inspected. Thus, the purpose of this paper is to describe a new program, TagQC, which allows a computerized QC GC analysis that identifies automatically, swiftly, and accurately suspicious TLD GC

Instrumental parameters' determination in a fluorescences X-ray Philips PW 1400 equipment

International Nuclear Information System (INIS)

Martinez, J.M.; Fasio, I.; Baronio, N.; Viola, M.

1987-01-01

The instrumental parameters of a Philips PW 1400 equipment wavelengths dispersive are determined; fundamentally, those related to the equipment's accuracy (stability at a very short, short and long term drift) as well as to those related to the detection system (dead time, detector's cleaning and detection limit). (S.M.) [es

Computerized photogrammetry used to calculate the brow position index.

Science.gov (United States)

Naif-de-Andrade, Naif Thadeu; Hochman, Bernardo; Naif-de-Andrade, Camila Zirlis; Ferreira, Lydia Masako

2012-10-01

The orbital region is of vital importance to facial expression. Brow ptosis, besides having an impact on facial harmony, is a sign of aging. Various surgical techniques have been developed to increase the efficacy of brow-lift surgery. However, no consensus method exists for an objective measurement of the eyebrow position due to the curvature of the face. Therefore, this study aimed to establish a method for measuring the eyebrow position using computerized photogrammetry. For this study, 20 orbital regions of 10 volunteers were measured by direct anthropometry using a digital caliper and by indirect anthropometry (computerized photogrammetry) using standardized digital photographs. Lines, points, and distances were defined based on the position of the anthropometric landmarks endocanthion and exocanthion and then used to calculate the brow position index (BPI). Statistical analysis was performed using Student's t test with a significance level of 5 %. The BPI values obtained by computerized photogrammetric measurements did not differ significantly from those obtained by direct anthropometric measurements (p > 0.05). The mean BPI was 84.89 ± 10.30 for the computerized photogrammetric measurements and 85.27 ± 10.67 for the direct anthropometric measurements. The BPI defined in this study and obtained by computerized photogrammetry is a reproducible and efficient method for measuring the eyebrow position. This journal requires that authors assign a level of evidence to each article.

Greta Kaits 9. apr. 1928 - 29. aug. 1946 : [luuletused] / Mats Traat

Index Scriptorium Estoniae

Traat, Mats, 1936-

2008-01-01

Sisu : Greta Kaits 9. apr. 1928 - 29. aug. 1946 ; Villem Hänilane 1898 - 1952 ; Peeter Pedak 1897 - 1953 ; Max Äidse 1903 - 1959 ; Lavinia Laidam ; Buchard Pruus ; Sergei Krivokopõtov ; Allan Luht 5. VII 1962 - 17. V 2007 ; Karl Saarmets 9. III 1924 - 14. IX 1981 ; Raimo Ritsik

Les pertes après récolte en Afrique : examen analytique et synthèse ...

International Development Research Centre (IDRC) Digital Library (Canada)

Partout en Afrique, les pertes après récolte le long de la chaîne alimentaire, de la ferme à la table, mettent en péril la sécurité alimentaire des agriculteurs dépourvus de ressources. La réduction des pertes après récolte peut aider à accroître la disponibilité, la qualité nutritionnelle et la salubrité des aliments, et à améliorer ...

Radiographic analysis of body composition by computerized axial tomography

International Nuclear Information System (INIS)

Heymsfield, S.B.

1986-01-01

Radiographic methods of evaluating body composition have been applied for over five decades. A marked improvement in this approach occurred in the mid-nineteen-seventies with the introduction of computerized axial tomography. High image contrast, cross-sectional imaging and rapid computerized data processing make this technique a sophisticated clinically applicable tool. (author)