Nuclear design report for Kori nuclear power plant unit 4 cycle 8

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Kyoon; Jung, Yil Sub; Kim, Si Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-07-01

This report presents nuclear design calculations for cycle 8 of Kori unit 4. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s 48 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 421 EFPD corresponding to a cycle burnup of 16950 MWD/MTU. (Author) 8 refs., 55 figs., 17 tabs.

Measurement of gamma ray flux within the containment building at the first unit of Kori nuclear power station

Energy Technology Data Exchange (ETDEWEB)

Kim, T. W.; Kim, K. D.; Yoon, C. H.; Han, J. M.; Hu, Y. H. [Korea Hydraulic and Nuclear Power Company, Taejon (Korea, Republic of)

2004-07-01

To evaluate gamma ray dose response of GM counter being used for monitoring of gamma ray field in nuclear power plants, gamma ray energy spectra and fluxes were obtained for three positions at the unit 1 of the Kori nuclear power station. By applying the response values of Eberline's E112B survey meter to the results, the doses represented on the survey meter were overestimated from 1.31 to 1.37 times when compared to the real doses for these three positions.

The trip status and the reduction countermeasure in Kori nuclear power plant unit 1 and 2

International Nuclear Information System (INIS)

Kim, Jung-Soo

1991-01-01

Nuclear power account for 36% of Korea's total electric capacity and provided over 50% of the net electric power supply by June 1991. These plants supply US with the cheapest and most stable electric supply available. However each units capacity is very large and a plant trip due to failure of a component or a human error has a great influence on the nations electric power supply and drastically decreases the reserve margin. This report will analyze the trip causes and measure the trip frequency from the first commercial operation of Kori unit 1 and 2 to the end of June 1991, reflect to the plant operation, management and facility modification, etc. This will minimize the number of trips or urgent power reductions and thus contribute to an increase in plant capacity factor and safety, and stabilize the electric power demand and supply. The safety and the economy of nuclear power plant have to be secured and raised respectably by increasing the capacity factor. Since the prevention of trips plays an important role in the plant safety and economy, we have to do our best to prevent the unexpected trip

IAEA Completes Expert Mission to Kori 1 Nuclear Power Plant in the Republic of Korea

International Nuclear Information System (INIS)

2012-01-01

Full text: An international team of nuclear safety experts led by the International Atomic Energy Agency (IAEA) has completed a review of safety practices at the Kori 1 Nuclear Power Plant (NPP) near Busan in the Republic of Korea. The IAEA assembled the team at the request of Korea Hydro and Nuclear Power Co., Ltd. (KHNP) following a station blackout event on 9 February 2012. The team - comprised of experts from Belgium, France, Sweden, United Kingdom and the IAEA - conducted its mission from 4 to 11 June 2012 under the leadership of the IAEA's Division of Nuclear Installation Safety. The expert mission applied the methodology of the IAEA's Operational Safety Review (OSART) missions and covered the areas of Management, Organization and Administration; Operations; Maintenance and Operating Experience. The conclusions of the review are based on the IAEA's Safety Standards, which are developed by the Agency to help nations improve their nuclear safety practices, which are the responsibility of every nation that undertakes nuclear-related activities. Throughout the review, the exchange of information between the experts and plant personnel was very open, professional and productive. Prior to the mission, Korea's Nuclear Safety and Security Commission completed an interim investigation, and it continues to perform additional investigations and technical reviews. The Commission identified corrective actions for the plant concerning reinforcing safety culture, emergency diesel generator reliability, configuration control and risk management during refueling outage, test and maintenance procedures and emergency action level declaration. The expert mission confirmed that some corrective actions have already been completed and others are in progress. The expert mission found the management and staff of Kori 1 NPP to be committed and working hard to complete all improvements. The root cause analysis of the event at Kori 1 NPP is still in progress and is expected to lead to

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.

Assessment of Coping Capability of KORI Unit 1 under Extended Loss AC Power and Loss of Ultimate Heat Sink Initiated by Beyond Design Natural Disaster

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hyun; Ha, Sang Jun [KHNP CRI, Daejeon (Korea, Republic of); Han, Kee Soo [Nuclear Engineering Service and Solution (NESS) Co. Ltd., Deajeon (Korea, Republic of); Park, Chan Eok [KEPCO Engineering and Constructd., Deajeon (Korea, Republic of)

2016-10-15

In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. This paper presents assessment results of coping capability of KORI Unit 1 under the simultaneous Extended Loss of AC Power (ELAP) and Loss of Ultimate Heat Sink (LUHS) which is a representative plant condition initiated by beyond design natural disaster. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings.

Assessment of Coping Capability of KORI Unit 1 under Extended Loss AC Power and Loss of Ultimate Heat Sink Initiated by Beyond Design Natural Disaster

International Nuclear Information System (INIS)

Kim, Chang Hyun; Ha, Sang Jun; Han, Kee Soo; Park, Chan Eok

2016-01-01

In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. This paper presents assessment results of coping capability of KORI Unit 1 under the simultaneous Extended Loss of AC Power (ELAP) and Loss of Ultimate Heat Sink (LUHS) which is a representative plant condition initiated by beyond design natural disaster. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings

Development of the Kori 1 simulator for the MCR modernization

Energy Technology Data Exchange (ETDEWEB)

Lee, Myeong Soo; Hong, Jin Hyuk [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Kori Unit 1 is the first commercial nuclear power reactor, pressurized water reactor that came into commercial service in April 1978 and is licensed for continued operation till 2017. The Main Control Board (MCB) was designed and was not applied the Human Factor Engineering (HFE) program during the construction phase but was performed the D CRDR (Detailed Control Room Design Review) as a post-TMI action. So Korea Hydro Nuclear Power, Ltd. (KHNP) has selected the hybrid type MCR by considering the existing equipment conditions and the operability of the plant as follows. .. Operator console upgrade, .. Plant Computer System (PCS) upgrade, .. PAS (Plant alarm System) upgrade, .. Remote Shutdown Panel (RSP) upgrade, .. Electrical control panel upgrade, and .. Interior improvement including lighting system. KINS will conduct the safety review of the new control room in Kori 1 as the same review level of construction permit (CP), operating license (OL) process. KHNP Central research Institute (CRI) developed a Kori 1 Full Scope Simulator (FSS) to have HFE Verification and Validation Test and operator training for the new modernized I and C of the MCR. This paper describes the several features and the results of the Kori 1 FSS. The NSSS thermal hydraulics model for the Kori 1 Simulator was developed by using the RELAP5 RT code, the real time version of RELAP5 developed by Idaho National Laboratory(INL). It could be configured from RELAP5/MOD3.2 by choosing the correct set of conditional coding and the base RELAP5 nodalization was shown in Figure 2. The NESTLE is a true two energy group neutronics code that computes the neutron flux and power for each node at every time step.

Development of radiation protection technology for application of the retired steam generator, Kori Unit no. 1

Energy Technology Data Exchange (ETDEWEB)

Kim, J. H.; Jang, D. C.; Song, K. S.; Lee, S. J.; Ahn, C. S.; Kim, D. H.; Im, Y. K.; Kim, H. D. [Hanil Nuclear Co., Ltd., Anyang (Korea, Republic of)

2005-04-15

It is a field study to develop and verify maintenance technologies such as verification and technology development of ECT (Eddy current test) using failure, heat tube excavation and field pressure test regarding the utilization of retired steam generator using 2 units of Retired Steam Generator in Kori 1 that was replaced for the first time in Korea in 1998. Since May, 2003, our team has investigated Retired Steam Generator which was stored in Radioactive waste warehouse in Korea Hydro and Nuclear Power Kori unit no.1 Branch, in order to study natural fault ECT signal acquisition, maintenance technology verification, small tubes/samples abstraction. A temporal task zone was made focusing on 'Man Way at the bottom of Chamber 'A'.' The purpose of the study is to establish Radiological Protection and Radioactive Waste Treatment Plan by setting ALARA (As Low As Reasonably Achievable) goal systematically, which is the basic concept of Radiological Protection and reduction in exposure of radiological workers to radioactive materials with proper Radiological Protection countermeasures according to the changes in radioactivity, to prevent expansion from contamination and to manage 'Radioactive Waste Reduction Activities' effectively.

Life extension program of KORI Unit 1 NPP in Korea

International Nuclear Information System (INIS)

Hong, Sun-Yull

1998-01-01

The two phases of Life extension program for KORI Unit 1 NPP are presented. Phase I is completed. It was concluded that life extension is a feasible option in technical and economic aspects. Detailed analysis of RPV is underway, plan for Phase II is finished. It deals with screening and sorting of all relevant SSCs, detailed life evaluation of SSCs, ageing management program and documentation for license renewal application

Reload safety evaluation report for Kori nuclear power unit 1, cycle 14

International Nuclear Information System (INIS)

Kim, Joo Young; Kim, Oh Hwan; Nam, Kee Il; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

1994-05-01

This report presents the reload safety evaluation for Kori-1, Cycle 14 and demonstrate that the reactor core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 14 core design described herein. (Author) 1 refs., 9 figs., 5 tabs

Fan Cooler Operation in Kori 1 for Mitigating Severe Accident

International Nuclear Information System (INIS)

Suh, Nam Duk; Park, Jae Hong

2005-01-01

The Korea Ministry of Science and Technology (MOST) issued the 'Policy on Severe Accident of Nuclear Power Plants' in August 2001. According to the policy it was required for the licensee to develop a plant specific severe accident management guideline (SAMG) and to implement it. Thus the utility has made an implementation plan to develop SAMGs for operating plants. The SAMG for Kori unit 1 was submitted to the government on January 2004. Since then, the government trusted KINS to review the submitted SAMG in view of its feasibility and effectiveness. The first principle of the developed SAMG is to use only the available facilities as it is without introducing any system change. Because Kori-1 has no mitigative facility against combustible gases during severe accident, it relies heavily both on spray and on fan cooler systems to control the containment condition. Thus one of the issues raised during the review is to know whether the fan coolers which are designed for DBA LOCA can be effective in mitigating the severe accident conditions. This paper presents an analysis result of fan cooler operation in controlling the containment condition during severe accident of Kori 1

Postirradiation examination of Kori-1 nuclear power plant fuels

International Nuclear Information System (INIS)

Ro, S.G.; Kim, E.K.; Lee, K.S.; Min, D.K.

1994-01-01

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institue. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity. (orig.)

Postirradiation examination of Kori-1 nuclear power plant fuels

Science.gov (United States)

Seung-Gy, Ro; Eun-Ka, Kim; Key-Soon, Lee; Duck-Kee, Min

1994-05-01

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institute. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity.

Marine ecosystem analysis for Kori nuclear power plant

International Nuclear Information System (INIS)

Lee, C.H.; Kim, Y.H.; Cho, T.S.

1980-01-01

The effect of both radioactive and thermal effluents discharged from the plant on aquatic ecosystem is one of the primary concerns in evaluating the environmental impact due to the operation of the nuclear power plant. Biological alteration of aquatic ecosystems may be resulted from radioactive effluents, thermal pollution and chemical releases. There is also another possible synergistic effect, that is, the combination of the above stresses, which may cause an impact severer than that of the sum of the individual impact. This report deals with species diversity and seasonal variations of those numbers of phytoplankton, marine algae and microorganisms, and distribution of radioactivity of marine organisms, as well as those data pertaining to sea water analysis. The present survey is designed to provide a partial baseline information for environmental impact assessment of Kori nuclear plant unit no. 1. (author)

Development of a GUI based RETRAN running environment for Kori NPP units 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Doo

2000-09-01

RETRAN was developed by EPRI and introduced for domestic use. RETRAN, which is a best-estimate system code approved by USNRC and used by most utilities in US, can be used in various plant support activities such as licensing calculations for plant design changes, EOP validation, and training. RETRAN, however, has been limited to only a few groups of specialists because of the difficulty involved in its usage. The aim of this project is to develop a graphic user interface (GUI) based code running environment for RETRAN named PRE (RETRAN Running Environment) in order to assist ordinary users in their input preparation, code execution, and output interpretation. TRIP and CONTROL BLOCK and VOLUME/JUNCTION input cards from base input are designed to be able to modify the existing input cards and add a new input cards through dialog boxes for users who have not much expertise in use of RETRAN. The RRE is designed to provide the calculated results though on-line X-Y graphs, plant mimics, indicators, nodalization window for easy interpretation of its output. It also provides the replay function using pre-calculated results saved in files. The RRE was developed for Kori NPP units 1 and 2 using Delphi 4.0 and Visual Fortran 6.0 and it runs on personal computers to increase the accessibility. The RRE developed in this study for Kori units 1 and 2 can be used in various plant support activities which require thermal-hydraulic analysis of the NSSS (Nuclear Steam Supply System) such as licensing calculations for plant design change, validation of EOP improvement, and operator training. The RRE developed can be expanded its application to other nuclear plants with low expense.

Development of a GUI based RETRAN running environment for Kori NPP units 1 and 2

International Nuclear Information System (INIS)

Kim, Kyung Doo

2000-09-01

RETRAN was developed by EPRI and introduced for domestic use. RETRAN, which is a best-estimate system code approved by USNRC and used by most utilities in US, can be used in various plant support activities such as licensing calculations for plant design changes, EOP validation, and training. RETRAN, however, has been limited to only a few groups of specialists because of the difficulty involved in its usage. The aim of this project is to develop a graphic user interface (GUI) based code running environment for RETRAN named PRE (RETRAN Running Environment) in order to assist ordinary users in their input preparation, code execution, and output interpretation. TRIP and CONTROL BLOCK and VOLUME/JUNCTION input cards from base input are designed to be able to modify the existing input cards and add a new input cards through dialog boxes for users who have not much expertise in use of RETRAN. The RRE is designed to provide the calculated results though on-line X-Y graphs, plant mimics, indicators, nodalization window for easy interpretation of its output. It also provides the replay function using pre-calculated results saved in files. The RRE was developed for Kori NPP units 1 and 2 using Delphi 4.0 and Visual Fortran 6.0 and it runs on personal computers to increase the accessibility. The RRE developed in this study for Kori units 1 and 2 can be used in various plant support activities which require thermal-hydraulic analysis of the NSSS (Nuclear Steam Supply System) such as licensing calculations for plant design change, validation of EOP improvement, and operator training. The RRE developed can be expanded its application to other nuclear plants with low expense

The experiences to improve plant performance and reliability of Ko-Ri nuclear power plants

International Nuclear Information System (INIS)

Kang, Ho Weon

1998-01-01

This paper provides a discussion of the lessons learned from operational experience and the future plans to improve performance of the Ko-Ri plant. To operate nuclear power plants safely with good performance is the only way to mitigate the negative image of nuclear power generation to the public and to enhance the economical benefit compared to other electrical generation method. Therefore, in a continuous effort to overcome a negative challenge from outside, we have driven an aggressive 'OCTF' campaign as part of safety. As a result of our efforts, the following remarkable achievements have been accomplished. (1) 3 times of OCTF during recent three years (2) Selected twice as a top notch power plant on the list of NEI magazine in terms of plant capacity factor (3) No scram recorded in 1997 for all 4 units at Ko-Ri site. Ko-Ri is now undergoing the large scale plant betterment projects for retaking-off our operating performance to the level of new challenge target. Such improvement of critical components in the reactor coolant system and turbine system greatly contribute to increase the safety and reliability of the plant and to shortening of the planned outage period as well as to reduction of radiation exposure and radwaste. (Cho, G. S.). 5 tabs., 10 figs

Reload safety evaluation report for Kori nuclear power plant unit 1, cycle 13

International Nuclear Information System (INIS)

Park, Chan Oh; Moon, Bok Ja; Cho, Byeong Ho; Nam, Kee Il; Kim, Oh Hwan; Chang, Doo Soo; Yoon, Han Young; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

1993-03-01

This report presents the reload safety evaluation for Kori-1, Cycle 13 and demonstrates that the reactor core being composed of various fuel assembly types applied in this evaluation will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 13 core and results are described in this report. (Author)

Development of the Real-time Core and Thermal-Hydraulic Models for Kori-1 Simulator

Energy Technology Data Exchange (ETDEWEB)

Hong, Jin Hyuk; Lee, Myeong Soo; Hwang, Do Hyun; Byon, Soo Jin [KEPRI, Daejeon (Korea, Republic of)

2010-10-15

The operation of the Kori-Unit 1 (1723.5MWt) is expanded to additional 10 years with upgrades of the Main Control Room (MCR). Therefore, the revision of the procedures, performance tests and works related with the exchange of the Main Control Board (MCB) are currently carried out. And as a part of it, the fullscope simulator for the Kori-1 is being developed for the purpose of the pre-operation and emergence response capability for the operators. The purpose of this paper is to report on the performance of the developed neutronics and thermal-hydraulic (TH) models of Kori Unit 1 simulator. The neutronics model is based on the NESTLE code and TH model based on the RELAP5/MOD3 thermal-hydraulics analysis code which was funded as FY-93 LDRD Project 7201 and is running on the commercial simulator environment tool (the 3KeyMaster{sup TM} of the WSC). As some examples for the verification of the developed neutronics and TH models, some figures are provided. The outputs of the developed neutronics and TH models are in accord with the Nuclear Design Report (NDR) and Final Safety Analysis Report (FSAR) of the reference plant

Determination of optimum pressurizer level for kori unit 1

Energy Technology Data Exchange (ETDEWEB)

Song, Dong Soo; Lee, Chang Sup; Yong, Lee Jae; Kim, Yo Han; Lee, Dong Hyuk [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1998-12-31

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are performed. The methodology is developed by evaluating {sup d}ecrease in secondary heat removal{sup e}vents such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling. The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. 6 refs., 5 figs. (Author)

Determination of optimum pressurizer level for kori unit 1

Energy Technology Data Exchange (ETDEWEB)

Song, Dong Soo; Lee, Chang Sup; Lee Jae Yong; Kim, Yo Han; Lee, Dong Hyuk [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1997-12-31

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are performed. The methodology is developed by evaluating {sup d}ecrease in secondary heat removal{sup e}vents such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling. The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. 6 refs., 5 figs. (Author)

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

Design Modification of Kori Unit 1 for the Equipment Qualification

International Nuclear Information System (INIS)

Park, J. W.; Kim, M. Y.; Han, K. T.; Park, J. D.

2007-01-01

There has not been a strict regulatory requirements for the Equipment Qualification(EQ) in 1970's when Kori Unit 1 had begun the construction and the commercial operation. The Korean regulatory body requested the EQ on the various safety-related components, as a result of Periodic Safety Review. However, the EQ itself is impossible in some areas, due to the high pressure/temperature and flooding environment conditions from the pipe breaks. Design modification is being considered in the Auxiliary Building, the Intermediate Building, the Component Cooling Water Heat Exchanger Building and the Turbine Building, in order to mitigate the environmental conditions for the EQ

The experiences to improve plant performance and reliability of Ko-Ri nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kang, Ho Weon [Korea Electric Power Corp. Ko-Ri nuclear power division, Ko-Ri (Korea, Republic of)

1998-07-01

This paper provides a discussion of the lessons learned from operational experience and the future plans to improve performance of the Ko-Ri plant. To operate nuclear power plants safely with good performance is the only way to mitigate the negative image of nuclear power generation to the public and to enhance the economical benefit compared to other electrical generation method. Therefore, in a continuous effort to overcome a negative challenge from outside, we have driven an aggressive 'OCTF' campaign as part of safety. As a result of our efforts, the following remarkable achievements have been accomplished. (1) 3 times of OCTF during recent three years (2) Selected twice as a top notch power plant on the list of NEI magazine in terms of plant capacity factor (3) No scram recorded in 1997 for all 4 units at Ko-Ri site. Ko-Ri is now undergoing the large scale plant betterment projects for retaking-off our operating performance to the level of new challenge target. Such improvement of critical components in the reactor coolant system and turbine system greatly contribute to increase the safety and reliability of the plant and to shortening of the planned outage period as well as to reduction of radiation exposure and radwaste. (Cho, G. S.). 5 tabs., 10 figs.

The 5th surveillance testing for Kori unit 2 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-03-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

Dose mapping in working space of KORI unit 1 using MCNPX code

Energy Technology Data Exchange (ETDEWEB)

Lee, C. W.; Shin, C. H.; Kim, J. G. [Hanyang University, Seoul (Korea, Republic of); Kim, S. Y. [Innovative Techonology Center for Radiation Safety, Seoul (Korea, Republic of)

2004-07-01

Radiation field analysis in nuclear power plant mainly depends on actual measurements. In this study, the analysis using computational calculation is performed to overcome the limits of measurement and provide the initial information for unfolding. The radiation field mapping is performed, which makes it possible to analyze the trends of the radiation filed for whole space. By using MCNPX code, containment building inside is modeled for KORI unit 1 cycle 21 under operation. Applying the neutron spectrum from the operating reactor as a radiation source, the ambient doses are calculated in the whole space, containment building inside, for neutron and photon fields. Dose mapping is performed for three spaces, 6{approx}20, 20{approx}44, 44{approx}70 ft from bottom of the containment building. The radiation distribution in dose maps shows the effects from structures and materials of components. With this dose maps, radiation field analysis contained the region near the detect position. The analysis and prediction are possible for radiation field from other radiation source or operating cycle.

Relative power density distribution calculations of the Kori unit 1 pressurized water reactor with full-scope explicit modeling of monte carlo simulation

International Nuclear Information System (INIS)

Kim, J. O.; Kim, J. K.

1997-01-01

Relative power density distributions of the Kori unit 1 pressurized water reactor calculated by Monte Carlo modeling with the MCNP code. The Kori unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starting with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system coverages to a κ value of 1.00039 ≥ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root mean square error of 2.159%. (author)

Upgrade of KNPEC no.2 Simulator for Kori Unit 3 Power Uprating

Energy Technology Data Exchange (ETDEWEB)

Hong, Jin-Hyuk; Lee, Seung-Ho [KEPRI, Daejeon (Korea, Republic of)

2007-07-01

Kori-Unit 3 and 4 is preparing the operation of the power-uprating (2900MWt), and therefore the Korea regulatory body(KINS) requested the operator training with the simulator reflecting the power-uprating. As a result of the intensive research and expertise of KEPRI on the simulators, KEPRI accomplished the upgrade project of KNPEC no.2 simulator for Kori-Unit 3 power-uprating. This project includes various high-tech methods incorporating - realtime neutronics model based on MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors) code, best-estimate neutronics code by the KINS, (By using the RMASTER, the precision of the simulation of the neutron behaviors in the core is highly improved.) - betterment of the reactor coolant system and the balance-of-plant system - modification of the corresponding setpoints due to the power-uprating And the acceptance test procedure (ATP) was successfully carried out through the integration of system models and its performance tests. Through the success of this project, the operator training for the power uprating of the Kori-Unit 3 will be accomplished before its power operation and, after all, this simulator will contribute to the safe operation for the power-uprating of the Kori-Unit 3 and 4.

Field measurement of the piping system vibration of Ko-Ri unit 4 during the load-following operation

International Nuclear Information System (INIS)

Chung, Tae-Young; Hong, Sung-Yull; Kim, Bum-Nyun.

1989-01-01

During the load-following operation of nuclear power plants, flow rate, temperature, and pressure in the piping system can be varied by changing the electric power output level, and these variations can cause different vibration phenomena in the piping system. The piping system vibration is important because it is directly related to the dynamic stress of the piping system and can affect the life of the piping system through structural fatigue. An assessment of vibration levels for the classes II and III piping systems of the Ko-Ri Unit 4950-MW nuclear power plant was performed according to the given pattern of the load-following operation to study its feasibility from the viewpoint of piping system vibration. The classes II and III piping system vibration of the Ko-Ri Unit 4 may not cause any potential problem under the given pattern of the load-following operation; however, it is recommended that long-term operation in the 85 to 95% power range be avoided as much as possible

Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

The 5th surveillance testing for Kori unit 1 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-08-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

Multi-dimensional analysis of the ECC behavior in the UPI plant Kori Unit 1

International Nuclear Information System (INIS)

Bae, Sungwon; Chung, Bub-Dong; Bang, Young Seok

2008-01-01

A multi-dimensional transient analysis during the LBLOCA of the Kori Unit 1 has been performed by using the MARS code. Based on 1-D nodalization of the Kori Unit 1, the reactor vessel nodalizations have been replaced by the multi-dimensional component. The multi-dimensional component for the reactor vessel is designed as 5 radial, 8 peripheral, and 21 vertical grids. It is assumed that the fuel assemblies are homogeneously distributed in inner 3 radial grids. The outer 1 radial grid region is modeled as the core bypass. The outer-model 1 radial grid is used for the downcomer region. The corresponding heat structures and fuels are modified to fit for the multi-dimensional reactor vessel model. The form drag coefficients for the upper plenum and the core have been designated as 0.6 and 9.39, respectively. The form drag coefficients for the radial and peripheral directions are assigned to the same on the assumption of homogeneous distribution of the flow obstacles. After obtaining the 102% power steady operation condition, cold leg LOCA simulation is performed during 400 second period. The multi-dimensional steady run results show no severe differences compared to the traditional 1-D nodalization results. After the ECC injection starts, a liquid pool is maintained at the upper plenum because the ECCS water can not overcome the upward gas flow that comes from the reactor core through the upper tie plate. The depth of ECCS water pool is predicted as about 20% of the total height from the upper tie plate and the center line of the hot leg pipe. At the vicinity region of the active ECCS show higher depth of liquid pool. The accumulated water flow rate passing the upper tie plate is calculated by the transient result. Much downward water flow is obtained at the outer-most region of upper plenum space. The downward flow dominant region is about 32.3% of the total upper tie plate area. The accumulated ECCS bypass ratio is predicted as 27.64% at 300 second. It is calculated

Thermal Analysis for Environmental Qualification of Kori Nuclear power plant unit 3 and 4

International Nuclear Information System (INIS)

Seo, Kwi Hyun; Byun, Choong Sup; Song, Dong Soo

2006-01-01

This paper shows the temperature profiles of safety related electrical equipment exposed to MSLB inside containment. It must be demonstrated that the LOCA qualification conditions exceed or are equivalent to the maximum calculated MSLB conditions. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPT-LT/028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. Power uprating code that is, COPATTA benchmarking study performed in six equipment at saturation temperature and surface temperature. Specially, thermal analysis carefully investigate that view point environmental qualification and NUREG- 0588 be mentioned in regard to safety-related heat sink it boundary condition or geometry information

Thermal Analysis for Environmental Qualification of Kori Nuclear power plant unit 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Seo, Kwi Hyun [ENERGEO Inc., Sungnam (Korea, Republic of); Byun, Choong Sup; Song, Dong Soo [KEPRI, Taejon (Korea, Republic of)

2006-07-01

This paper shows the temperature profiles of safety related electrical equipment exposed to MSLB inside containment. It must be demonstrated that the LOCA qualification conditions exceed or are equivalent to the maximum calculated MSLB conditions. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPT-LT/028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. Power uprating code that is, COPATTA benchmarking study performed in six equipment at saturation temperature and surface temperature. Specially, thermal analysis carefully investigate that view point environmental qualification and NUREG- 0588 be mentioned in regard to safety-related heat sink it boundary condition or geometry information.

Neutron spectrum measurement inside containment vessel at Kori nuclear power plant unit 1

Energy Technology Data Exchange (ETDEWEB)

Han, J. M.; Kim, T. W.; Kim, K. D.; Youn, C. H. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

2003-10-01

There would be a case for the radiation worker have to work inside of the containment vessel to inspect or repair reactor facilities. In this case, the information about distribution of neutron field is needed to estimate neutron exposure dose of worker. Neutron spectra were measured by BMS(Bonner Multisphere Spectrometer) at 4 points of 6 ft and 20 ft, 2 points of 44 ft, 5 points of 70 ft in containment vessel of Kori unit 1. From the calculation, the following results were obtained. Neutron fluxes of 6 ft were between 2.623 x 10{sup 2} and 2.746 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 9.209 x 10{sup -6} and 3.377 x 10{sup -2} MeV, equivalent doses of neutron were between 0.025 and 2.675 mSv/hr. Neutron fluxes of 20 ft were between 1.771 x 10{sup 1} and 1.682 x 10{sup 3} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 6.084 x 10{sup -6} and 2.988 x 10{sup -1} MeV, equivalent doses of neutron were between 0.004 and 0.228 mSv/hr. Neutron fluxes of 44 ft were between 3.367 x 10{sup 2} and 3.483 x 10{sup 2} neutron / cm{sup 2}{center_dot}sec, average neutron energies were between 3.962 x 10{sup -2} and 7.360 x 10{sup -2} MeV, equivalent doses of neutron were between 0.069 and 0.089 mSv/hr. Neutron fluxes of 70 ft were between 4.553 x 10{sup 3} and 1.407 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 3.668 x 10{sup -4} and 6.764 x 10{sup -2} MeV, equivalent doses of neutron were between 0.449 and 2.660 mSv/hr.

Nuclear design report for Yonggwang nuclear power plant unit 1, cycle 8

International Nuclear Information System (INIS)

Cho, Young Chul; Kim, Jae Hak; Park, Sang Yoon; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan

1993-10-01

This report presents nuclear design calculations for cycle 8 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA's enriched by nominally 3.70 w/o U 235 . Among the KOFA's, 56 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 447 EFPD corresponding to a cycle burnup of 18020 MWD/MTU. (Author) 8 refs., 39 figs., 17 tabs

Nuclear design report for Yonggwang nuclear power plant unit 1, cycle 8

Energy Technology Data Exchange (ETDEWEB)

Cho, Young Chul; Kim, Jae Hak; Park, Sang Yoon; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-10-01

This report presents nuclear design calculations for cycle 8 of Kori unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 56 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 8 amounts to 447 EFPD corresponding to a cycle burnup of 18020 MWD/MTU. (Author) 8 refs., 39 figs., 17 tabs.

Containment Response Analysis for Equipment Qualification of Kori Nuclear Power Plant Unit 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Byun, Choong Sup; Song, Dong Soo; Hwang, Yong Jun [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc., Sungnam (Korea, Republic of)

2006-07-01

Equipment that is used to perform a necessary safety function must be capable of maintaining functional operability under all service condition postulated to occur during the installed life for the time it is required. The pressure and temperature analyses for loss of coolant accident and main steam line break accident provide the bounding test envelope inside containment for the operability evaluation of safety equipment in harsh environmental. This paper describes the results of the containment pressure and temperature analysis for the equipment qualification (EQ) envelopes of Kori unit 3 and 4.

Reevaluation of Kori Unit 4 Natural Circulation Test

Energy Technology Data Exchange (ETDEWEB)

Yassin, Nassir [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Woo, Sweng Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

The simulation results showed that the natural circulation flow developed by density difference was capable of removing decay heat from the fuel rod. The maximum pellet centerline temperature of the hot channel showed large margin to the pellet melting temperature. The maximum coolant temperature in the hot channel was well below the saturation temperature. If steam generators provide heat sink to the primary coolant system and thus natural circulation is maintained, the integrity of the fuel in the core can be sustained with large margin. Passive cooling of reactor is inevitable in case of failures in forced cooling system such as loss of electric power for cooling pumps. Fukushima accident showed the importance of the passive core cooling. During the commissioning test of PWRs, natural circulation test is performed to demonstrate the passive core cooling by natural convection. The driving force for coolant flow is developed by the density deference along the loop multiplied by the gravitation. Using the data from 'natural circulation test' and 'RCS flow coast down test' of Kori Unit 4, fuel behavior was reevaluated by FRAPTRAN code. RCS natural circulation test of Kori Unit 4 was reevaluated by FRAPTYRAN simulation to study the fuel behavior during the flow coast down transient and at the equilibrium condition in which decay heat transport and RCS flow were stabilized.

A Preliminary Study on the Containment Integrity following BIT Removal for Kori NPP Unit 3,4

Energy Technology Data Exchange (ETDEWEB)

Song, Dong Soo; Byun, Choong Sup [KEPRI, Nuclear Power Generation Laboratory, Daejeon (Korea, Republic of); Jo, Jong Young [ENERGEO Inc., Sungnam (Korea, Republic of)

2008-05-15

The Boron Injection Tank (BIT) is to provide high concentrated boric acid to the reactor in order to mitigate the consequences of postulated Main Steam Line Break accidents (MSLB). Although BIT plays an important role in mitigating the accident, high concentration of 20,000ppm causes valve leakage, pipe clog, precipitation and continuous heat tracing have to be provided. This paper is for the feasibility study of containment integrity using CONTEMPT code for BIT removal of Kori Nuclear Power Plant (NPP) Unit 3, 4.

A Preliminary Study on the Containment Integrity following BIT Removal for Kori NPP Unit 3,4

International Nuclear Information System (INIS)

Song, Dong Soo; Byun, Choong Sup; Jo, Jong Young

2008-01-01

The Boron Injection Tank (BIT) is to provide high concentrated boric acid to the reactor in order to mitigate the consequences of postulated Main Steam Line Break accidents (MSLB). Although BIT plays an important role in mitigating the accident, high concentration of 20,000ppm causes valve leakage, pipe clog, precipitation and continuous heat tracing have to be provided. This paper is for the feasibility study of containment integrity using CONTEMPT code for BIT removal of Kori Nuclear Power Plant (NPP) Unit 3, 4

Development of a Real-Time Thermal Performance Diagnostic Monitoring system Using Self-Organizing Neural Network for Kori-2 Nuclear Power Unit

International Nuclear Information System (INIS)

Kang, Hyun Gook; Seong, Poong Hyun

1996-01-01

In this work, a PC-based thermal performance monitoring system is developed for the nuclear power plants. the system performs real-time thermal performance monitoring and diagnosis during plant operation. Specifically, a prototype for the Kori-2 nuclear power unit is developed and examined is very difficult because the system structure is highly complex and the components are very much inter-related. In this study, some major diagnostic performance parameters are selected in order to represent the thermal cycle effectively and to reduce the computing time. The Fuzzy ARTMAP, a self-organizing neural network, is used to recognize the characteristic pattern change of the performance parameters in abnormal situation. By examination, the algorithm is shown to be ale to detect abnormality and to identify the fault component or the change of system operation condition successfully. For the convenience of operators, a graphical user interface is also constructed in this work. 5 figs., 3 tabs., 11 refs. (Author)

Thermal recovery characteristics of Kori-unit 1 linde 80 weld

International Nuclear Information System (INIS)

Chi, S. H.; Hong, J. H.; Kuk, I. H.; Kim, I. S.

1997-01-01

The recovery activation energy, the order of reaction and the characteristic recovery rate constant were determined by isochronal (573K -823K) and isothermal (723K - 775K) annealing experiments on specimens made from a broken half of a Kori-Unit 1 surveillance weld specimen (fluence: 1.21 x 10 23 n/m 2 , E (1MeV, Cu: 0.29 wt%) to investigate the recovery characteristics of a high copper weld of neutron-irradiated reactor pressure vessel (RPV). Vickers microhardness tests were conducted to trace the recovery behavior after heat treatments. The results were analyzed in terms of recovery stages, behavior of responsible defects and recovery kinetics. It was shown that recovery occurred through two annealing stages (stage I: 673K - 753K, stage II: 753K - 823K) with recovery activation energies of 2.68 eV and 2.83 eV for stage I and II, respectively. The isothermal hardness recovery at 723 K and 775 K coincided with the ratio of the characteristic rate constant for each recovery stage. The order of reaction was 2 for both recovery stages. The recovery activation energies of present specimens are approximately equal to that of copper diffusion in α-iron in the presence of vacancies, suggesting that recovery may occur through the diffusion of copper atoms. The present results strongly support the copper precipitate coarsening model. (author)

Numerical Simulation of Groundwater Flow at Kori Nuclear Power Plant Site

International Nuclear Information System (INIS)

Sohn, Wook; Sohn, Soon Whan; Chon, Chul Min; Kim, Kue Youn

2010-01-01

Recently, the understanding of hydrogeological characteristics of nuclear power sites is getting more importance with increasing public concerns over the environment since such understanding is essential for an environmentally friendly operation of plants. For such understanding, the prediction of groundwater flow pattern onsite plays the most critical role since it is the most dynamic of the factors to be considered. In this study, the groundwater flow at the Kori Plant 1 site has been simulated numerically with aim of providing fundamental information needed for improving the understanding of the hydrogeological characteristics of the site

The application experience of ethanol amine at KORI unit 4

International Nuclear Information System (INIS)

Lee, Sang-hak; Park, Jong-il; Lee, Jae-won; Kim, Guem-Soo

2004-01-01

The secondary system water chemistry in the KORI PWR units has been well controlled by reducing the corrosion in the tubes of the steam generators; the pH agent has been changing from ammonia to ethanol amine (ETA) (1.8 to 2.0 ppm). For example, the iron concentration in the system was reduced by 40 to 70% compared with ammonia treatment. The paper presents the detailed information, such as ETA injection concentration and the variation of pH, concentration of organic acid products, and irons in the system, and a significant change on sludge after ETA application. (S. Ohno)

Study on application of operating experience to new nuclear power plant

International Nuclear Information System (INIS)

Hong, Nam Pyo

1991-01-01

From the standpoint of designing the nuclear power plant, nine operating units have been designed and constructed as turn-key base by foreign Nuclear Steam Supply System (NSSS) Suppliers or as component base by foreign Architect/Engineer companies. In case of the component base project, the owner of electric company generally has merits that owner's operational experiences can be effectively incorporated from the beginning stage of design by A/E. Even though six nuclear units, Kori Units 3 and 4, Yonggwang Units 1 and 2, and Ulchin Units 1 and 2, were designed as component base by foreign A/E's, operational experience feedback from Kori Unit 1, such as design improvement and system upgrade, could not be reflected, because the design process of the following units started well ahead before Kori Unit 1 operating experience is obtained enough to reflect on future nuclear power plant design. It can be stated that foreign A/E's used their experience in designing nuclear projects on very limited basis

Continuous operation of NPP Kori Unit 1 - Fireproof paint for cables

International Nuclear Information System (INIS)

Wendt, Dipl-Ing. Ruediger; Kim, Duill; Sik, Cho Hong

2008-01-01

Fireproof cable coating materials have been used in European NPP, especially in Germany, Russia, Ukraine, Czech Republic, Lithuania, Switzerland. Wide experiences were made during operation while applying these systems. In NPP Kori, Unit 1, a fire proof cable coating project was realised for the first time in a NPP of KHNP. The scope of services of the cable trays to coat amounts to 15,587m 2 . In different fire compartments and rooms the cables should be coated partially respectively completely with the fire proof cable coating system. The extent of cable surfaces to coat was stipulated by KHNP on the basis of an analysis made by KHNP. The project was tendered on the basis of a technical specification of KHNP. The specification is mainly predicted on Korean and US standards. The most important criteria for the fire proof cable coating is resumed as follows: The fireproof cable coating has to assure the fire protection of the cables for a period defined and for operational conditions defined in such a manner that the general conditions for the operation of the cable installation will not be affected

Review of the research proposal for the steam generator retired from Kori unit 1

Energy Technology Data Exchange (ETDEWEB)

Kim, Joung Soo; Han, Joung Ho; Kim, Hong Pyo; Lim, Yun Soo; Lee, Deok Hyun; Hwang, Seong Sik; Hur, Do Haeng [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

The tubes of the steam generator retired form Kori unit 1 have many different kinds of failures, such as denting pitting, wastage, ODSCC, PWSCC.Korea Electric Power Research Institute (KEPRI) submitted a research proposal for the steam generator to the Korea Institute S and T Evaluation and Planning (KSITEP). The KISTEP requested Korea Atomic Energy Research Institute to review the proposal by organizing a committee which should be composed of the specialists of the related domestic research institutes. Opinions of the committee on the objectives, research fields, economic benefit and validity in the research proposal were reviewed and suggested optimal research fields to be fulfilled successfully for the retired steam generator. Also, the rolls for the participants in the research works were allocated, which is critical in order to do the project effectively. 6 figs., 5 tabs. (Author)

Marine-ecosystem analysis for the Kori nuclear power plant

International Nuclear Information System (INIS)

Lee, J.H.; Kim, Y.H.

1979-01-01

The effects of radioactive effluents and warm water discharged from the plant on aquatic ecosystem is one of the primary considerations in evaluating the impact due to the operation of the nuclear power plant. Biological alteration of aquatic ecosystems may be resulted from radioactive effluents, thermal pollution and chemical releases; there is also the possible synergistic effect, that is, the combination of the above stresses, which may cause an effect greater than that of the sum of the individual effects. This report deals with species diversity and seasonal vegetation of phytoplankton, marine algae and microorganisms, radioactive contamination of marine organisms, and lateral distribution of sea water temperature from discharge point. The present investigation is designed to provide a partial baseline information for environmental safety against Kori nuclear power plant. (author)

Analysis of a station blackout transient at the Kori units 3/4

International Nuclear Information System (INIS)

Bang, Young Seok; Kim, Hho Jung

1992-01-01

A transient analysis of station blackout accident is performed to evaluate the plant specific capability to cope with the accident at the Kori Units 3/4. The RELAP5/MOD3/5m5 code and full three loop modelling scheme are used in the calculation. The leak flow from reactor coolant system due to a failure of reactor coolant pump seal following the accident is assumed to be 25 gpm and the turbine driven aux feedwater unavailable. As a result, it is found that no core uncovery occurs in the plant until 7100 sec following a station blackout, the steam generator has a decay heat removal capability until 3100 sec, and the natural circulation over the reactor coolant loop until the complete loop seal voiding are observed. And the Nuclear Plant Analyzer is used and found to be effective in improving the phenomenological understanding on the accident

Reload safety evaluation report for kori nuclear power plant unit 4, cycle 8

International Nuclear Information System (INIS)

Park, Chan Oh; Jung, Yil Sup; Kim, Si Yong; Kim, Ki Hang; Kwon, Hyuk Sung; Oh, Dong Seok; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

1993-06-01

This report presents the reload safety evaluation for Kori-4, Cycle 8 and demonstrate that the reactor core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licening bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 8 core design described herein. (Author)

Nuclear power plant life time improvement and management program in Korea

International Nuclear Information System (INIS)

Sung Yull Hong; Ill Seok Jeong; Taek Ho Song

1995-01-01

Korea Electric Power Research Institute (KEPRI) of Korea Electric Power Corporation (KEPCO) has performed a lifetime management of nuclear power plant program (LMNPP), ''Nuclear Power Plant Lifetime Management (PLIM) (I)'', since November 1993, which is a feasibility study of the Kori Unit 1 lifetime management including aging evaluation of the thirteen major components. The results of the PLIM(I) will provide information which is necessary for decision making of the Kori Unit 1 lifetime improvement. A plan of the work scope and schedule for the next phase, PLIM(II), will also be provided by this project. This paper introduced KEPRI's basic strategy of LMNPP, PLIM organization, current status, some interim results of the PLIM(I), and other related programs in Korea. So far, we have done field data survey, systems/structures screening, components prioritization, lifetime evaluation methodology study, and fracture mechanics tests of the Kori Unit 1 reactor pressure vessel surveillance coupons. Currently life assessment of the major components and PLIM economic evaluation of Kori Unit 1 are under way. (author)

Present status of nuclear containments in Korea

International Nuclear Information System (INIS)

Park, Jihong; Hong, Jaekeun; Lee, Byunghoon; Son, Youngho

2007-01-01

Since the first nuclear power plant in Korea, Kori unit no.1, was started in commercial service in 1978, 20 units including Kori unit no.1 have been operated and maintained until now in Korea. Recently several units were started to be constructed and also, additionally more than 4 units were planned to be constructed in the near future. The importance of nuclear containments has been always one of the hottest issues for the safety and protection of nuclear power plants until now. At the beginning of nuclear power plants construction in Korea, several typed nuclear containment systems were adopted. For those reasons, various codes, standards, and inspection technologies are applied to nuclear containment systems differently. In this study, the status of inservice inspection performed for the safety and maintenance of nuclear containments in Korea was researched. Overall nuclear containment systems and inspections performed up to recently in Korea including trends, inspection items, periods, and regulations were described briefly. (author)

Cutset Quantification Error Evaluation for Shin-Kori 1 and 2 PSA model

International Nuclear Information System (INIS)

Choi, Jong Soo

2009-01-01

Probabilistic safety assessments (PSA) for nuclear power plants (NPPs) are based on the minimal cut set (MCS) quantification method. In PSAs, the risk and importance measures are computed from a cutset equation mainly by using approximations. The conservatism of the approximations is also a source of quantification uncertainty. In this paper, exact MCS quantification methods which are based on the 'sum of disjoint products (SDP)' logic and Inclusion-exclusion formula are applied and the conservatism of the MCS quantification results in Shin-Kori 1 and 2 PSA is evaluated

The application of FIMS to the Korean nuclear power plants

International Nuclear Information System (INIS)

Lee, Seung-Tae; Choi, Dae-Sik; Yang, Choong-Yeol

2010-01-01

The smart transmitters have been installed at selected systems of Ulchin Nuclear Units 5 and 6, and diagnosis and calibration for the smart transmitters have been carried out using the mobile calibration tools. In Shin-Kori Nuclear Units 1 and 2 and Shin-Wolsong Nuclear Units 1 and 2, which are under design and construction, the smart transmitters are installed in the places where total integrated radiation dose is below 10 Gy. FIMS (Field Instrument Management System) is applied to Shin-Kori Nuclear Units 1 and 2 and Shin-Wolsong Nuclear Units 1 and 2 to monitor and calibrate the smart transmitters remotely. FIMS is a centralized management system to allow the users to monitor the operating status, to perform the instrument diagnosis, and to carry out the calibration of locally mounted transmitters remotely during the commissioning and normal operation. (authors)

System and Software Design for the Plant Protection System for Shin-Hanul Nuclear Power Plant Units 1 and 2

International Nuclear Information System (INIS)

Hwang, In Seok; Kim, Young Geul; Choi, Woong Seock; Sohn, Se Do

2015-01-01

The Reactor Protection System(RPS) protects the core fuel design limits and reactor coolant system pressure boundary for Anticipated Operational Occurrences (AOOs), and provides assistance in mitigating the consequences of Postulated Accidents (PAs). The ESFAS sends the initiation signals to Engineered Safety Feature - Component Control System (ESF-CCS) to mitigate consequences of design basis events. The Common Q platform Programmable Logic Controller (PLC) was used for Shin-Wolsung Nuclear Power Plant Units 1 and 2 and Shin-Kori Nuclear Power Plant Units 1, 2, 3 and 4 since Digital Plant Protection System (DPPS) based on Common Q PLC was applied for Ulchin Nuclear Power Plant Units 5 and 6. The PPS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) was developed using POSAFE-Q PLC for the first time for the PPS. The SHN1 and 2 PPS was delivered to the sites after completion of Man Machine Interface System Integrated System Test (MMIS-IST). The SHN1 and 2 PPS was developed to have the redundancy in each channel and to use the benefits of POSAFE-Q PLC, such as diagnostic and data communication. The PPS application software was developed using ISODE to minimize development time and human errors, and to improve software quality, productivity, and reusability

System and Software Design for the Plant Protection System for Shin-Hanul Nuclear Power Plant Units 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Hwang, In Seok; Kim, Young Geul; Choi, Woong Seock; Sohn, Se Do [KEPCO EnC, Daejeon (Korea, Republic of)

2015-10-15

The Reactor Protection System(RPS) protects the core fuel design limits and reactor coolant system pressure boundary for Anticipated Operational Occurrences (AOOs), and provides assistance in mitigating the consequences of Postulated Accidents (PAs). The ESFAS sends the initiation signals to Engineered Safety Feature - Component Control System (ESF-CCS) to mitigate consequences of design basis events. The Common Q platform Programmable Logic Controller (PLC) was used for Shin-Wolsung Nuclear Power Plant Units 1 and 2 and Shin-Kori Nuclear Power Plant Units 1, 2, 3 and 4 since Digital Plant Protection System (DPPS) based on Common Q PLC was applied for Ulchin Nuclear Power Plant Units 5 and 6. The PPS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) was developed using POSAFE-Q PLC for the first time for the PPS. The SHN1 and 2 PPS was delivered to the sites after completion of Man Machine Interface System Integrated System Test (MMIS-IST). The SHN1 and 2 PPS was developed to have the redundancy in each channel and to use the benefits of POSAFE-Q PLC, such as diagnostic and data communication. The PPS application software was developed using ISODE to minimize development time and human errors, and to improve software quality, productivity, and reusability.

Development of Level-2 PSA Technology: A Development of the Database of the Parametric Source Term for Kori Unit 1 Using the MAAP4 Code

Energy Technology Data Exchange (ETDEWEB)

Kang, Chang Soon; Mun, Ju Hyun; Yun, Jeong Ick; Cho, Young Hoo; Kim, Chong Uk [Seoul National University, Seoul (Korea, Republic of)

1997-07-15

To quantify the severe accident source term of the parametric model method, the uncertainty of the parameters should be analyzed. Generally, to analyze the uncertainties, the cumulative distribution functions(CDF`S) of the parameters are derived. This report introduces a method of derivation of the CDF`s of the basic parameters, FCOR, FVES and FDCH. The calculation tool of the source term is the MAAP version 4.0. In the MAAP code, there are model parameters to consider an uncertain physical and/or chemical phenomenon. In general, the parameters have not a point value but a range. In this paper, considering this point, the input values of model parameters influencing each parameter are sampled using LHS. Then, the calculation results are shown in the cumulative distribution form. For a case study, the CDF`s of FCOR, FVES and FDCH of KORI unit 1 are derived. The target scenarios for the calculation are the ones whose initial events are large LOCA, small LOCA and transient, respectively. It is found that the distributions of this study are consistent to those of NUREG-1150 and are proven to be adequate in assessing the uncertainties in the severe accident source term of KORI Unit 1. 15 refs., 27 tabs., 4 figs. (author)

Assessment on the Reactor Containment Cooling Capability of Kori Unit 1 Under LOCA Conditions with Loss of Offsite Power

International Nuclear Information System (INIS)

Lee, Jin Yong; Park, Jong Woon; Kim, Hyeong Taek

2006-01-01

The fan cooler system is designed to remove heat from containment under postulated accident conditions. During a postulated LOCA concurrent with a Loss of Offsite Power (LOOP), the Component Cooling Water (CCW) pumps that supply cooling water to the fan cooler and the fan that supplies containment air to the fan cooler will temporarily lose power. Then, the high temperature steam in the containment atmosphere will pass over the fan cooler tubing without forced cooling water flow. In that case, boiling may occur in the fan cooler tubes causing steam bubbles to form and pass into the attached CCW piping creating steam voids. Prior to the CCW pumps restart, the presence of steam and subcooled water can induce the potential for water hammer. As the CCW pumps restart, the accumulated steam condenses and the pumped water can produce a water hammer when the void closes. The hydrodynamic loads caused by such a water hammer event could challenge the integrity and the function of the fan cooler and associated CCW system. With respect to this phenomena, the United States Nuclear Regulatory Commission (USNRC) issued the Generic Letter (GL) 96-06, which requests an assessment of the possibility of boiling and water hammer in the cooling water system. The objectives of this study are to develop a analysis method for predicting the thermal hydraulic status of containment fan cooler and then to assess the containment fan cooler of Kori Unit 1 using the developed model under a LOCA with LOOP

Analysis of effects on plant performance by major measuring points in the secondary systems of Kori nuclear power plant units 3 and 4

International Nuclear Information System (INIS)

Lee, Jung Woon; Park, Jae Chang; Lee, Jung Woon; Kim, Jung Taek; Chang, Soon Heung; Lee, Gwang Gu; Heo, Gyun Young; Lee, Sung Jin; Han, Kyu Hyun; Shin, Byung Soo

2003-06-01

In this study, correlation analysis was achieved for the major sensor position and the behavior of secondary system in Kori NPP unit 3, 4. Using the data from simulation model, the correlation between sensor position and electrical output, the correlation between sensor position and heat rate, and the correlation between different sensor positions were analyzed. On the basis of study results, a performance evaluation model was proposed, which can carry out secondary system performance diagnosis

Application of a Virtual Ovation System to the ShinKori-3 Simulator

International Nuclear Information System (INIS)

Hong, Jin Hyuk; Lee, Myeong Soo; Chung, Kyung Hun

2011-01-01

The Ovation system for the Shin-Kori Unit 3 Simulator is essentially a non-redundant, repackaged subset of the actual plant I and C equipment, with additional interface computers (SimStations). This system also present in the simulator to provide communication between the plant model computer and the stimulated Ovation equipment. The stimulated Ovation equipment in the simulator system includes Operator HMI (Human- Machine Interface) equipment and Ovation virtual controllers hosted by Virtual Controller Host workstations, which are not present in the actual plant DCS system. The simulator for the Shin-Kori Unit 3 and 4 is being developed by Korea Hydro and Nuclear Power's Central Research Institute (KHNP CRI). One of the features of the simulator is its application of a virtual Ovation system capable of simulated functionalities such as run, freeze, snapshot, backtrack and others required by ANSI/ANS-3.5 in addition to the original functionality for the actual Ovation system applied at the plant. This is the first application of a virtual Ovation system to a full-scope simulator for a nuclear power plant in Korea. The purpose of this paper is to provide the overall architecture of the communication system between the virtual system and the simulator model and to describe the current situation of the development of the system and recent relevant studies

Relaxation of inservice test frequency requirement for Kori 1 ASME code pumps

International Nuclear Information System (INIS)

Sohn, Gap Heon; Choi, Hae Yoon; Min, Kyung Sung; Rim, Nam Jin

1994-08-01

The objective of this investigation is to evaluate the technical and regulational requirements to justify the relaxation of the test frequency of Kori 1 pumps through reviewing the related rules and codes and standards, technical specifications of Kori 1 and other similar plants, standard technical specifications, research results for tech. spec. improvements and site test records. It is concluded that the relaxation of test frequency to quarterly be justified based on the conformance with rules and codes and standard, quarterly test cases in similar plants and standard tech. spec., recommendations of research result and stable site test records. (Author) 16 refs., 26 figs., 13 tabs

Sludge Removal and Retrieval of Foreign Object in SG of Kori Nuclear Power Plant, Unit 4

International Nuclear Information System (INIS)

Jeong, Wootae; Kim, Sangtae; Kim, Youngkug; Kang, Seokchul

2014-01-01

Sludge deposit was removed and foreign objects were inspected and retrieved on secondary side tube sheet of the SG during January 23 and February 22, April 15 and 27 in 2013. FOLAS-I lancing system, video probe and retrieval tools were used for lancing and foreign object removal respectively. Operators of the lancing system participated in mock-up training before doing the service to minimize operation time and radiation dose. Foreign objects were searched on top of 7 th TSP (tube support plate), on annulus and in tube bundle. Four objects were found and removed on annulus and in tube bundle. During the 21 st OH of Kori NPP unit 4, we removed 345.9 kilo gram of sludge and four foreign objects from three steam generators. Foreign objects which were removed from inside of SG showed us that relatively large object such as the hooked bolt might exists in steam generators. We can conclude that identifying and removing foreign object is very important to avoid possible tube failure. Removing circular metal of 152.4 gram also was successfully removed

Development of Severe Accident Management Strategies for Shin-Kori 3 and 4

International Nuclear Information System (INIS)

Lee, Youngseung; Kim, Hyeongtaek; Shin, Jungmin

2013-01-01

Shin-Kori units 3 and 4 are new reactors under construction as an APR 1400 type reactor. The plants which considered coping with severe accident from design phase are different from other operating plants in view of severe accident management strategies. The purpose of this paper is to establish optimal strategies for Shin-Kori 3 and 4. A scheme for optimized severe accident management was drawn up with the object of achieving core cooling, containment integrity, and decreased release of fission product. Shin-Kori units 3 and 4 are a new reactor and designed to add mitigating systems for coping with severe accident such as ECSBS, PAR, and CFS. Also the plants are reflected as a part of Fukushima followup measures The strategies of SAMG for Shin-Kori 3 and 4 were developed. The strategic approach was based on the concept of defense in depth. Firstly, strategies for core cooling were chosen such as RCS depressurization, injection to SG, injection to RCS, and injection to reactor cavity. Secondly, the plans for containment integrity were developed for controlling pressure and hydrogen in containment. Lastly, reduced release of fission product was considered for protection of the public after containment failure. The achieved strategies meet the needs of effective methods for severe accident management and enhancement of safety

Radiographic monitoring of the ossification of long bones in kori (Ardeotis kori) and white-bellied (Eupodotis senegalensis) bustards

International Nuclear Information System (INIS)

Naldo, J.L.; Samour, J.H.; Bailey, T.A.

1998-01-01

A serial radiographic study was conducted on eight kori bustard (Ardeotis kori) and four white-bellied bustard (Eupodotis senegalensis) chicks to determine the pattern of long bone development and to establish radiographic standards for assessing skeletal maturity. The ossification pattern, appearance of secondary ossification centres, and epiphyseal fusion of the long bones in kori and white-bellied bustards were similar to those in houbara bustards (Chlamydotis undulata macqueenii),rufous-crested bustards (Eupodotis ruficrista), domestic fowl (Gallusgallus), house wrens (Troglodytes aedon aedon), racing pigeons (Columba livia), and barn owls (Tyto alba). Secondary ossification centres were present at the proximal and distal tibiotarsus, proximal tarsometatarsus and proximal metacarpal III. The ossification of long bones occurred earlier in female kori bustards compared with males

Pretreatment Process for performance Improvement of SIES at Kori Unit 2 in Korea

International Nuclear Information System (INIS)

Lee, Sang Jin; Yang, Ho Yeon; Shin, Sang Woon; Song, Myung Jae

1994-01-01

Pretreatment process consisted of submerged hollow-fiber microfiltration(HMF) membrane and spiral-wound nanofiltration(SNF) membrane has been developed by NETEC, KHNP for the purpose of improving the impurities of liquid radioactive waste before entering Selective Ion Exchange System(SIES). The lab-scale combined system was installed at Kori NPP no. 2 nuclear power plant and demonstration tests using actual liquid radioactive waste were carried out to verify the performance of the combined system. The submerged HMF membrane was adopted for removal of suspended solid in liquid radioactive waste and the SNF membrane was used for removal of particulate radioisotope such as, Ag-110m and oily waste because ion exchange resin can not remove particulate radioisotopes. The liquid waste in Waste Holdup Tank(WHT) was processed with HMF and SNF membrane, and SIES. The initial SS concentration and total activity of actual waste were 38,000ppb and 1.534x10 -3 μCi/cc, respectively. The SS concentration and total activity of permeate were 30ppb and lower than LLD(Lower Limit of Detection), respectively

Estimation of residual stress distribution for pressurizer nozzle of Kori nuclear power plant considering safe end

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-08-15

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which considered safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

Evaluation on Cooling Performance of Containment Fan Cooler during Design Basis Accident with Loss of Offsite Power for Kori 3 and 4 Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Sung Bok; Lee, Sang Won [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Park, Young Chan [Atomic Creative Technology Co., LTD., Daejeon (Korea, Republic of)

2007-10-15

The purpose of this study is to evaluate cooling performance of containment fan cooler units and to review a technical background related to Generic Letter 96-06. In case that design basis accident (DBA) and loss of offsite power (LOOP) occurs, component cooling water (CCW) pumps cannot provide the cooling water source to fan cooler units while fan coolers coast down. Fan cooler units and CCW pumps are restarted by emergency diesel generator (EDG) operation and it takes about 30 seconds. In this scenario, before the EDG restarts and CCW flowrate is restored, heated air in the containment passes through coil of fan cooler units without cooling water source. In this situation, the boiling of water in the fan cooler units may occur. Restarting of CCW pumps may bring about condensation by injected cooling water and water hammer may occur. This thermal-hydraulic effect is sensitive to system configuration, i.e system pressure, containment pressure/temperature, EDG restarting time, etc. In this study, the evaluation of containment fan cooler units was performed for Kori 3 and 4 nuclear power plant.

ALARA review of the maintenance and repair jobs of repetitive high radiation dose at Kori Unit 3 and 4

International Nuclear Information System (INIS)

Cho, Y.H.; Moon, J.H.; Kang, C.S.; Lee, J.S.; Lee, D.H.

2003-01-01

The policy of maintaining occupational radiation dose (ORD) as low as reasonably achievable (ALARA) requires the effective reduction of ORD in the phases of design as well as operation of nuclear power plants. It has been identified that a predominant portion of ORD arises during maintenance and repair operations at nuclear power plants. The cost-effective reduction of ORD cannot be achieved without a comprehensive analysis of accumulated ORD data of existing nuclear power plants. To identify the jobs of repetitive high ORD, the ORD data of Kori Units 3 and 4 over 10-year period from 1986 to 1995 were compiled into the PC-based ORD database program. As the radiation job classification structure, 26 main jobs are considered, most of which are further subdivided into detailed jobs. According to the order of the collective dose values for 26 main jobs, 10 jobs of high collective dose are identified. As an ALARA review, then, top 10 jobs of high collective dose are statistically analyzed with regard to 1) dose rate, 2) crew number and 3) job frequency that are the factors determining the collective dose for the radiation job of interest. Through the ALARA review, main reasons causing to high collective dose values are identified as follows. The high collective dose of RCP maintenance job is mainly due to the large crew number and the high job frequency. The characteristics of refueling job are similar to those of RCP maintenance job. However, the high collective doses of SG-related jobs such as S/G nozzle dam job, S/G man-way job and S/G tube maintenance job are mainly due to high radiation dose rate. (author)

Fracture Toughness Evaluation of Kori-1 RPV Beltline Weld for a Long-Term Operation

International Nuclear Information System (INIS)

Lee, Bong-Sang; Kim, Min-Chul; Ahn, Sang-Bok; Kim, Byung-Chul; Hong, Jun-Hwa

2007-01-01

Irradiation embrittlement of RPV (reactor pressure vessel) material is the most important aging issue for a long-term operation of nuclear power plants. KORI unit 1, which is the first PWR in Korea, is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties have a certain level of the safety margin. The current regulation requires Charpy V-notch impact data through conventional surveillance tests. It is based on the assumption that Charpy impact test results are well correlated with the fracture toughness properties of many engineering steels. However, Charpy V-notch impact data may not be adequate to estimate the fracture toughness of certain materials, such as Linde 80 welds. During the last decade, a tremendous number of fracture toughness data on many RPV steels have been produced in accordance with the new standard test method, the so-called master curve method. ASTM E1921 represents a revolutionary advance in characterizing fracture toughness of RPV steels, since it permits establishing the ductile to brittle transition portion of the fracture toughness curve with direct measurements on a relatively small number of relatively small specimens, such as pre-cracked Charpy specimens. Actual fracture toughness data from many different RPV steels revealed that the Charpy test estimations are generally conservative with the exception of a few cases. Recent regulation codes in USA permit the master curve fracture toughness methodology in evaluating an irradiation embrittlement of commercial nuclear reactor vessels

Analysis of inadvertent safety injection incident at Kori unit 3 on september 6, 1990

International Nuclear Information System (INIS)

Kim, Kyun Tae; Chung, Bub Dong; Kim, In Goo; Kim, Hho Jung

1992-01-01

The inadvertent safety injection incident occurred at Kori Unit 3 on September 6, 1990 is analyzed using RELAP5/MOD3 code. The event was initiated by a failure of main feedwater control valve in one of three steam generators. The actual sequence of plant transient with the proper estimations of the operator actions is investigated in the present calculation. The calculational results are compared with the plant transient data. It is shown that the results of the plant behaviors are in good agreement with the plant data. The emergency response guidelines is assessed for the time of the SI termination and the establishment of natural circulation. The changes in the time of the SI termination do not significantly affect the overall plant behaviors, and the natural circulation is established

Optimal replacement and inspection periods of safety and control boards in Wolsung nuclear power plant unit 1

International Nuclear Information System (INIS)

Mok, Jin Il

1993-02-01

In nuclear power plants, the safety and control systems are important for operating and maintaining safety of nuclear power plants. Due to the failure of the instrument and control devices of nuclear power plants caused by aging, nuclear power plants occasionally trip. Since the start of first commercial operation of Kori nuclear power plant (NPP) unit 1, the trips caused by instrument and control systems account for 28% of total trips of NPPs in Korea. Even a single trip of a nuclear power plant causes an extravagant economical loss and deteriorates public acceptance of nuclear power plants. Therefore, the replacement of the instrument and control devices with proper consideration of the aging effect is necessary in order to prevent the inadvertent trip. In this work we investigated the optimal replacement periods of the digital control computer's (DCC) and the programmable digital comparator's (PDC) electronic circuit boards of Wolsung nuclear power plant Unit 1. We first derived mathematical models which calculate optimal replacement periods for electronic circuit boards of digital control computer (DCC) and for those of the programmable digital comparator (PDC) in Wolsung NPP unit 1. And we analytically obtained the optimal replacement periods of electronic circuit boards by using these models. We compared these periods with the replacement periods currently used at Wolsung NPP Unit. The periods used at Wolsung is not based on mathematical analysis, but on empirical knowledge. As a consequence, the optimal replacement periods analytically obtained for the electronic circuit boards of DCC and those used in the field shown small difference : the optimal replacement periods analytically obtained for the electronic circuit boards of PDC are shorter than those used in the field in general. The engineered safeguards of Wolsung nuclear power plant unit 1 contains redundant systems of 2-out-of-3 logic which are not operating under normal conditions but they are called

Effects of environmental radiation of Kori nuclear power plant on the human population

International Nuclear Information System (INIS)

Kim, Y.J.

1979-01-01

In order to clarify and protect the effects of environmental radiation according to the operation of Kori nuclear power plant on human population, the base line survey for the human monitoring, the fauna of land nocturnal insects, and the karyotypes of amphibian species which have been living around the power plant site were carried out. ''Kilchunri'' population which took for the human monitoring lie within a 2km distance from power plant site. Human monitoring, house and food characteristics, individual experience of X-ray exposures, human chromosome analysis and fauna of nocturnal land insects were surveyed and expressed in numerical tables. Chromosome number obtained from the amphibia which were collected around the power plant area was as follows; Kaloula borealis 2N=30, Rana amurensis 2N=26, Rana dybouskii 2N=24, Rana rugosa 2N=26, Rana migromaculata 2N=26, Rana plancyi 2N=26, Bombina orientalis 2N=24, Hyla arborea 2N=24, Bufo stejnegeri 2N=22, and Bufo bufo 2N=22. (author)

76 FR 40754 - Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units...

Science.gov (United States)

2011-07-11

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0100; Docket Nos. 50-413 and 50-414; Docket Nos. 50-369 and 50-370; Docket Nos. 50-269, 50-270, And 50-287] Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3; Notice...

Advanced Passive Reactors : Leading The U. S. Nuclear Renaissance

International Nuclear Information System (INIS)

Henderson, Ronald R.

1990-01-01

Twenty-one years have passed since Korea Electric Power Corporation and Westinghouse announced plans to build Kori 1. Today, Korea's nuclear program is one of the most successful in the world. The electricity generated from Kori 1 and eight other nuclear plants has helped to spark the remarkable growth and transformation of Korea into a modern industrial power. Westinghouse is proud to have been Korea's partner on six of those plants. It the past is the bast prophet of the future, then you and your countrymen should certainly be excited by your future. Korean industry is poised to continue its steady growth, and that means continued growth for your nuclear industry. Currently, the U. S. nuclear industry is experiencing a similar mood of excitement. In fact, it would be necessary to go almost all the way back to the beginning of the birth of the Korean nuclear industry, in 1969, to find a time when the future of nuclear power in the United States looked as bright as it does today. Part of our excitement stems from the welcome prospect of growth. In recent years, there has not been a market for new nuclear plants in the United States. Utilities either had excess capacity or were building plants they had ordered before 1974. For example, between 1980 and 1989, U. S. utilities completed 46 large nuclear units, but didn't order a single new one in that time. Since 1983, however, strong economic growth in the United States has caused the demand for electric power to grow about twice as fast as utilities had projected. Today, utilities will need to order new busload plants. When they do, utilities won't want technology developed 20 years ago. They'll be looking for plants that can meet the environmental, economic, and safety standards of the 21st century

Effect of preemptive weld overlay sequence on residual stress distribution for dissimilar metal weld of Kori nuclear power plant pressurizer

Energy Technology Data Exchange (ETDEWEB)

Bae, Hong Yeol; Song, Tae Kwang; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea Univ., Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-07-01

Weld overlay is one of the residual stress mitigation method which arrest crack. An overlay weld sued in this manner is termed a Preemptive Weld OverLay(PWOL). PWOL was good for distribution of residual stress of Dissimilar Metal Weld(DMW) by previous research. Because range of overlay welding is wide relatively, residual stress distribution on PWR is affected by welding sequence. In order to examine the effect of welding sequence, PWOL was applied to a specific DMW of KORI nuclear power plant by finite element analysis method. As a result, the welding direction that from nozzle to pipe is better good for residual stress distribution on PWR.

Effect of preemptive weld overlay sequence on residual stress distribution for dissimilar metal weld of Kori nuclear power plant pressurizer

International Nuclear Information System (INIS)

Bae, Hong Yeol; Song, Tae Kwang; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae; Lee, Kyoung Soo; Park, Chi Yong

2008-01-01

Weld overlay is one of the residual stress mitigation method which arrest crack. An overlay weld sued in this manner is termed a Preemptive Weld OverLay(PWOL). PWOL was good for distribution of residual stress of Dissimilar Metal Weld(DMW) by previous research. Because range of overlay welding is wide relatively, residual stress distribution on PWR is affected by welding sequence. In order to examine the effect of welding sequence, PWOL was applied to a specific DMW of KORI nuclear power plant by finite element analysis method. As a result, the welding direction that from nozzle to pipe is better good for residual stress distribution on PWR

Safety R and D Needs in Multi-Unit PSA from a Regulatory Viewpoint

Energy Technology Data Exchange (ETDEWEB)

Heo, Gyunyoung; Yoon, Ji Woong [Kyung Hee University, Yongin (Korea, Republic of); Kim, Man Cheol [Chung-Ang University, Seoul (Korea, Republic of)

2016-10-15

The Nuclear Safety and Security Commission (NSSC) NSSC’s recent approval of the construction permits and operation licenses of Shin Kori units 3 to 6 has brought attention to the multi-unit probabilistic safety assessment (PSA). The disaster in Fukushima revealed the risks associated with the use of spent fuel pools (SFPs), and the linear (or non-linear) consequences of radiation leaks from multi-units in particular, has raised concerns. The number of nuclear power plants (NPPs) in operation will increase to ten at the Kori and Shin Kori sites when the construction of Units 5 and 6 is complete - the highest concentration of power reactors in one region in world history. Nuclear disasters do not happen frequently, but when one does happen, its cascading effects can inflict massive human and environmental loss. Proper safety measures are needed as more nuclear installations are put in place. Nuclear regulators are forced to promptly produce a sufficient degree of realistic and objective risk assessment results to meet the rapidly-increasing demand for multi-unit PSA, while striving to achieve matching levels of technological development. The issues with the technological aspects of PSA would still be inherent even after a few years of research, and ensuring the safety of multi-units would still prove challenging. Experts are however relieved that the general public is now much more familiar with the concept of PSA than in the past, which indicates a rise in the degree of public understanding of technology. More studies should be done on multi-unit regulation and safety so as to steer the government in the proper direction for enacting policies aimed to ensure the health of local residents. In conclusion, the technical problems associated with multi-unit PSA that were not mentioned in this paper are essentially similar to their single-unit counterparts, therefore, reinforcing the base infrastructure of PSAs in general will work towards resolving such technical

Locating the Source of Atmospheric Contamination Based on Data From the Kori Field Tracer Experiment

Directory of Open Access Journals (Sweden)

Piotr Kopka

2015-01-01

Full Text Available Accidental releases of hazardous material into the atmosphere pose high risks to human health and the environment. Thus it would be valuable to develop an emergency reaction system which can recognize the probable location of the source based only on concentrations of the released substance as reported by a network of sensors. We apply a methodology combining Bayesian inference with Sequential Monte Carlo (SMC methods to the problem of locating the source of an atmospheric contaminant. The input data for this algorithm are the concentrations of a given substance gathered continuously in time. We employ this algorithm to locating a contamination source using data from a field tracer experiment covering the Kori nuclear site and conducted in May 2001. We use the second-order Closure Integrated PUFF Model (SCIPUFF of atmospheric dispersion as the forward model to predict concentrations at the sensors' locations. We demonstrate that the source of continuous contamination may be successfully located even in the very complicated, hilly terrain surrounding the Kori nuclear site. (original abstract

Science, society, and America's nuclear waste: Unit 1, Nuclear waste

International Nuclear Information System (INIS)

1992-01-01

This is unit 1 in a four-unit secondary curriculum. It is intended to provide information about scientific and societal issues related to the management of spent nuclear fuel from generation of electricity at nuclear powerplants and high-level radioactive waste from US national defense activities. The curriculum, supporting classroom activities, and teaching materials present a brief discussion of energy and electricity generation, including that produced at nuclear powerplants; information on sources, amounts, location, and characteristics of spent nuclear fuel and high-level radioactive waste; sources, types and effects of radiation; US policy for managing and disposing of spent nuclear fuel and high-level radioactive waste and what other countries are doing; and the components of the nuclear waste management system

The estimated evacuation time for the emergency planning zone of the Kori nuclear site, with a focus on the precautionary action zone

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Hee; Jeong, Jae Jun [School of Mechanical Engineering, Pusan National University, Busan (Korea, Republic of); Shin, Won Ki; Song, Eun Young; Cho, Cheol Woo [Div. of Nuclear Safety, Busan Metropolitan City, Busan (Korea, Republic of)

2016-09-15

The emergency planning zone (EPZ) of the city of Busan is divided into the precautionary actions zone (PAZ) and the urgent protective action planning zone; which have a 5-km radius and a 20-km to 21-km radius from the nuclear power plant site, respectively. In this study, we assumed that a severe accident occurred at Shin-Kori nuclear unit 3 and evaluated the dispersion speed of radiological material at each distance at various wind speeds, and estimated the effective dose equivalent and the evacuation time of PAZ residents with the goal of supporting off-site emergency action planning for the nuclear site. The total effective dose equivalent, which shows the effect of released radioactive materials on the residents, was evaluated using the RASCAL 4.2 program. In addition, a survey of 1,036 residents was performed using a standardized questionnaire, and the resident evacuation time according to road and distance was analyzed using the VISSIM 6.0 program. According to the results obtained using the VISSIM and RASCAL programs, it would take approximately 80 to 252.2 minutes for permanent residents to move out of the PAZ boundary, 40 to 197.2 minutes for students, 60 to 232.2 minutes for the infirm, such as elderly people and those in a nursing home or hospital, and 30 to 182.2 minutes for those temporarily within the area. Consequently, in the event of any delay in the evacuation, it is estimated that the residents would be exposed to up to 10 mSv·h-1 of radiation at the Exclusion Area Boundaries (EAB) boundary and 4-6 mSv·h-1 at the PAZ boundary. It was shown that the evacuation time for the residents is adequate in light of the time lapse from the initial moment of a severe accident to the radiation release. However, in order to minimize the evacuation time, it is necessary to maintain a system of close collaboration to avoid traffic congestion and spontaneous evacuation attempts.

The estimated evacuation time for the emergency planning zone of the Kori nuclear site, with a focus on the precautionary action zone

International Nuclear Information System (INIS)

Lee, Jang Hee; Jeong, Jae Jun; Shin, Won Ki; Song, Eun Young; Cho, Cheol Woo

2016-01-01

The emergency planning zone (EPZ) of the city of Busan is divided into the precautionary actions zone (PAZ) and the urgent protective action planning zone; which have a 5-km radius and a 20-km to 21-km radius from the nuclear power plant site, respectively. In this study, we assumed that a severe accident occurred at Shin-Kori nuclear unit 3 and evaluated the dispersion speed of radiological material at each distance at various wind speeds, and estimated the effective dose equivalent and the evacuation time of PAZ residents with the goal of supporting off-site emergency action planning for the nuclear site. The total effective dose equivalent, which shows the effect of released radioactive materials on the residents, was evaluated using the RASCAL 4.2 program. In addition, a survey of 1,036 residents was performed using a standardized questionnaire, and the resident evacuation time according to road and distance was analyzed using the VISSIM 6.0 program. According to the results obtained using the VISSIM and RASCAL programs, it would take approximately 80 to 252.2 minutes for permanent residents to move out of the PAZ boundary, 40 to 197.2 minutes for students, 60 to 232.2 minutes for the infirm, such as elderly people and those in a nursing home or hospital, and 30 to 182.2 minutes for those temporarily within the area. Consequently, in the event of any delay in the evacuation, it is estimated that the residents would be exposed to up to 10 mSv·h-1 of radiation at the Exclusion Area Boundaries (EAB) boundary and 4-6 mSv·h-1 at the PAZ boundary. It was shown that the evacuation time for the residents is adequate in light of the time lapse from the initial moment of a severe accident to the radiation release. However, in order to minimize the evacuation time, it is necessary to maintain a system of close collaboration to avoid traffic congestion and spontaneous evacuation attempts

Final report on effects of environmental radiation of Kori nuclear power plant on human population

International Nuclear Information System (INIS)

Kim, Y.J.; Kim, J.B.; Chung, K.H.; Lee, K.S.; Kim, S.R.; Yang, S.Y.

1980-01-01

In order to clarify and protect the effects of environmental radiation according to the operation of Kori nuclear power plant on the human population, the base line survey for the human monitoring, human life habits, expected individual exposure dose, frequencies of chromosomal aberration, gene frequencies and karyotypes in amphibia, fauna, and radiation sensitivities in microorganisms which have been living around the power plant site were carried out. Kilchonri population which took for the human monitoring lie within a 2 km distance from the power plant site. Human monitoring, house and food characteristics, individual experience of x-ray exposures, human chromosome analysis and fauna were surveyed and expressed in numerical tables. Chromosome number obtained from the amphibia which were collected around the power plant area was as follows: Kaloula borealis 2N=30, Rana amurensis 2N=26, Rana dybouskii 2N=24, Rana rugosa 2N=26, Rana nigromaculata 2N=26, Rana plancyi 2N=26, Bombina orientalis 2N=24, Hyla arborea 2N=24, Bufo stejnegeri 2N=22, Bufo bufo 2N=22. (author)

HRD System and Experience in the Korean Nuclear Industry

Energy Technology Data Exchange (ETDEWEB)

Kang, Byoungkook [KHNP Nuclear Power Education Institute, Busan (Korea, Republic of)

2012-03-15

Korea began to nurture its nuclear energy pioneers in the 1950s when the government dispatched personnel in research and policy-making areas to foreign institutions. Then in 1959, KAERI was established and now plays a leading role in nuclear technology R and D. In addition, Korea's first research reactor, TRIGA Mark-II, was built and put into operation in 1962. This paved the way for advancements in operation and technical development of nuclear reactors. In turn, these accomplishments led to the birth of Korea's first commercial reactor, Kori Unit 1, in the 1970s, and HRD in the nuclear industry was put on the right track. However, the Korean nuclear industry remained heavily dependent on nuclear exporting countries such as the US, Canada, and France. Already confident in construction, Korea took the lead in building Kori Units 3 and 4 and Ulchin Units 1 and 2 in the 1980s, but the country was still in need of technological self-reliance. In order to achieve this, Korea proactively launched systematic HRD programs and dispatched nuclear professionals to overseas nuclear facilities to secure individuals competent in the areas of NPP operations, plant design, and major equipment manufacturing. Thanks to its diligent endeavors, Korea's nuclear entities established independent nuclear training institutes in the 1990s and began producing a large number of competent personnel. This allowed the country to ensure not only the best operation and maintenance engineers but also the essential nuclear technology required for plant design and equipment manufacturing. Since the beginning of the 21{sup st} century, Korea has been producing its nuclear personnel on its own and exchanging nuclear training instructors and trainees with other organizations in fields where specialized knowledge is needed. Furthermore, Korea is taking comprehensive nuclear HRD measures in response to the rising demand for human resources that result from ongoing construction of NPPs in

HRD System and Experience in the Korean Nuclear Industry

International Nuclear Information System (INIS)

Kang, Byoungkook

2012-01-01

Korea began to nurture its nuclear energy pioneers in the 1950s when the government dispatched personnel in research and policy-making areas to foreign institutions. Then in 1959, KAERI was established and now plays a leading role in nuclear technology R and D. In addition, Korea's first research reactor, TRIGA Mark-II, was built and put into operation in 1962. This paved the way for advancements in operation and technical development of nuclear reactors. In turn, these accomplishments led to the birth of Korea's first commercial reactor, Kori Unit 1, in the 1970s, and HRD in the nuclear industry was put on the right track. However, the Korean nuclear industry remained heavily dependent on nuclear exporting countries such as the US, Canada, and France. Already confident in construction, Korea took the lead in building Kori Units 3 and 4 and Ulchin Units 1 and 2 in the 1980s, but the country was still in need of technological self-reliance. In order to achieve this, Korea proactively launched systematic HRD programs and dispatched nuclear professionals to overseas nuclear facilities to secure individuals competent in the areas of NPP operations, plant design, and major equipment manufacturing. Thanks to its diligent endeavors, Korea's nuclear entities established independent nuclear training institutes in the 1990s and began producing a large number of competent personnel. This allowed the country to ensure not only the best operation and maintenance engineers but also the essential nuclear technology required for plant design and equipment manufacturing. Since the beginning of the 21 st century, Korea has been producing its nuclear personnel on its own and exchanging nuclear training instructors and trainees with other organizations in fields where specialized knowledge is needed. Furthermore, Korea is taking comprehensive nuclear HRD measures in response to the rising demand for human resources that result from ongoing construction of NPPs in Korea and the UAE

Accidental safety analysis methodology development in decommission of the nuclear facility

Energy Technology Data Exchange (ETDEWEB)

Park, G. H.; Hwang, J. H.; Jae, M. S.; Seong, J. H.; Shin, S. H.; Cheong, S. J.; Pae, J. H.; Ang, G. R.; Lee, J. U. [Seoul National Univ., Seoul (Korea, Republic of)

2002-03-15

Decontamination and Decommissioning (D and D) of a nuclear reactor cost about 20% of construction expense and production of nuclear wastes during decommissioning makes environmental issues. Decommissioning of a nuclear reactor in Korea is in a just beginning stage, lacking clear standards and regulations for decommissioning. This work accident safety analysis in decommissioning of the nuclear facility can be a solid ground for the standards and regulations. For source term analysis for Kori-1 reactor vessel, MCNP/ORIGEN calculation methodology was applied. The activity of each important nuclide in the vessel was estimated at a time after 2008, the year Kori-1 plant is supposed to be decommissioned. And a methodology for risk analysis assessment in decommissioning was developed.

The current status and future prospects of the Korean nuclear power industry

International Nuclear Information System (INIS)

Lee, J. J.

2006-01-01

Recently, countries all over the world are becoming aware of the values and importance of nuclear energy which can help respond to energy crises caused by a sharp rise in oil prices and protect the earth from global warming. Since 1978, when Kori Unit 1(587MW), opened the nuclear generation era as a semi-domestic energy resource in Korea which is absolutely in short supply of energy, nuclear power generation in Korea has developed continuously for the past 28 years. Four new units including the Yonggwang 5 and 6 and Ulchin 5 and 6 have been successfully completed, raising the total nuclear installed capacity to 17,716MW from 20 units. At present, the nuclear generation in Korea is stably supplying about 40% of total electric generation, which is the fundamental energy of the nation, supporting the dynamic economic growth of Korea. In particular, Korean nuclear industry has been achieving excellent performance in nuclear power plant operation. The average capacity factor in 2005 hit the record of 95.5%, surpassing the previous record of 94.2% in 2003 in two years. Kori Unit 4 and another four units were listed at the top five in the capacity factor rating list of 2005 released by Nucleonics Week. In 2005, the site for radioactive waste disposal, which had been a long-cherished hope and the largest pending issue of the nuclear industry, was successfully selected in Korea through resident ballot as the first case of a national policy project, and as such, a national agenda was solved after 19 long years. Such a method in site selection has a significant meaning and establishes an excellent precedence; a large national policy project was decided upon by the residents themselves. As one of the model countries of building and operating nuclear power plants and technological independence, Korea is willing to contribute to the common goals of the world nuclear circle which can be summarized into energy security and environment preservation, by sharing accumulated

76 FR 24538 - Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station...

Science.gov (United States)

2011-05-02

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-413 and 50-414; NRC-2011-0100; Docket Nos. 50-369 and 50-370; Docket Nos. 50-269, 50-270, and 50-287] Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3...

Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

Status of Korean nuclear industry and Romania-Korea cooperation in nuclear field

International Nuclear Information System (INIS)

Myung-Key, Lee

2005-01-01

Current status of electric power in Korea is characterized by the end of August 2004 by a total installed capacity of about 62,000 MW while the total electricity generation is about 342,000 GWh. The installed capacity of nuclear power is 17,716 MW, sharing 29% of total installed capacity and presenting 38% of total electricity generation in Korea. In accordance with the provisions of the Long Term Energy Plan during the past 40 years, the installed capacity in Korea has been drastically increased. In the 1960's, major sources of electricity generation were locally-mined anthracite coal and hydro, but in the 1970's it was the imported oil. However, through diversification policy the dependence of the imported oil has been rapidly reduced and the share of coal , gas and nuclear generation has been steadily increased. According to the long-term power development plan updated last year, which is extended to 2017, the installed capacity in the year 2017 will be about 88,000 MW. At that time nuclear power will become the largest, sharing 30% and the shares of coal and gas fired power will be steadily decreased. Concerning the Nuclear Power Projects, there are four different nuclear power sites along the coast of Korean peninsula, Yonggwang, Kori, Wolsong and Ulchin. In addition to the currently operating 20 nuclear power plants, there are 6 more nuclear power plants under construction at Shin-Wolsong and Shin-Kori sites. Our efforts to enhance the technology, economy and safety of the nuclear power plants will be continuously pursued. Wolsong unclear power units 1, 2, 3 and 4 are CANDU type reactors which are same type as Romanian Cernavoda nuclear power plants. Operational performance, in terms of capacity factor of NPPs, has remained well above the world average and recorded 91.4% last year. Also, last year, the frequency of unplanned trip was 0.6 time per reactor. In 2004, ten NPPs achieved OCTF, which stands for One Cycle Trouble Free in 2004. Wolsong unit 1

Prevention of nuclear fuel cladding materials corrosion

International Nuclear Information System (INIS)

Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

1983-01-01

The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

Science, Society, and America's Nuclear Waste: Nuclear Waste, Unit 1. Teacher Guide. Second Edition.

Science.gov (United States)

Department of Energy, Washington, DC. Office of Civilian Radioactive Waste Management, Washington, DC.

This guide is Unit 1 of the four-part series Science, Society, and America's Nuclear Waste produced by the U.S. Department of Energy's Office of Civilian Radioactive Waste Management. The goal of this unit is to help students establish the relevance of the topic of nuclear waste to their everyday lives and activities. Particular attention is…

Fuqing nuclear power of nuclear steam turbine generating unit No.1 at the implementation and feedback

International Nuclear Information System (INIS)

Cao Yuhua; Xiao Bo; He Liu; Huang Min

2014-01-01

The article introduces the Fuqing nuclear power of nuclear steam turbine generating unit no.l purpose, range of experience, experiment preparation, implementation, feedback and response. Turn of nuclear steam turbo-generator set flush, using the main reactor coolant pump and regulator of the heat generated by the electric heating element and the total heat capacity in secondary circuit of reactor coolant system (steam generator secondary side) of saturated steam turbine rushed to 1500 RPM, Fuqing nuclear power of nuclear steam turbine generating unit no.1 implementation of the performance of the inspection of steam turbine and its auxiliary system, through the test problems found in the clean up in time, the nuclear steam sweep turn smooth realization has accumulated experience. At the same time, Fuqing nuclear power of nuclear steam turbine generating unit no.1 at turn is half speed steam turbine generator non-nuclear turn at the first, with its smooth realization of other nuclear power steam turbine generator set in the field of non-nuclear turn play a reference role. (authors)

76 FR 1469 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

Science.gov (United States)

2011-01-10

... Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Environmental Assessment... Plant, LLC, the licensee, for operation of the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2... Impact Statement for License Renewal of Nuclear Plants, Calvert Cliffs Nuclear Power Plant (NUREG-1437...

78 FR 77726 - Exelon Generation Company, LLC Three Mile Island Nuclear Station, Unit 1

Science.gov (United States)

2013-12-24

... Three Mile Island Nuclear Station, Unit 1 AGENCY: Nuclear Regulatory Commission. ACTION: Exemption... License No. DPR-50, which authorizes operation of the Three Mile Island Nuclear Station, Unit 1 (TMI-1... Facility Operating License No. DPR-50, which authorizes operation of the Three Mile Island Nuclear Station...

75 FR 36700 - Exelon Generation Company, LLC; Three Mile Island Nuclear Station, Unit 1; Environmental...

Science.gov (United States)

2010-06-28

...; Three Mile Island Nuclear Station, Unit 1; Environmental Assessment and Finding of No Significant Impact... Company, LLC (the licensee), for operation of Three Mile Island Nuclear Station, Unit 1 (TMI-1), located... Three Mile Island Nuclear Station, Units 1 and 2, NUREG-0552, dated December 1972, and Generic...

75 FR 66802 - Calvert Cliffs Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

Science.gov (United States)

2010-10-29

... Nuclear Power Plant, LLC; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2; Notice of Withdrawal of...) has granted the request of Calvert Cliffs Nuclear Power Plant, LLC, the licensee, to withdraw its... for the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, located in Calvert County, MD. The...

Evaluation of Watts Bar Nuclear Plant Unit 1 Technical Specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Watts Bar Nuclear Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumption of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Watts Bar T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Watts Bar Nuclear Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Evaluation of Shoreham Nuclear Power Station, Unit 1 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Shoreham Nuclear Power Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Shoreham T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Shoreham Nuclear Power Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Second periodic safety review of Angra Nuclear Power Station, unit 1

Energy Technology Data Exchange (ETDEWEB)

Martins, Carlos F.O.; Crepaldi, Roberto; Freire, Enio M., E-mail: ottoncf@tecnatom.com.br, E-mail: emfreire46@gmail.com, E-mail: robcrepaldi@hotmail.com [Tecnatom do Brasil Engenharia e Servicos Ltda, Rio de Janeiro, RJ (Brazil); Campello, Sergio A., E-mail: sacampe@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

2015-07-01

This paper describes the second Periodic Safety Review (PSR2-A1) of Angra Nuclear Power Station, Unit 1, prepared by Eletrobras Eletronuclear S.A. and Tecnatom do Brasil Engenharia e Servicos Ltda., during Jul.2013-Aug.2014, covering the period of 2004-2013. The site, in Angra dos Reis-RJ, Brazil, comprises: Unit 1, (640 MWe, Westinghouse PWR, operating), Unit 2 (1300 MWe, KWU/Areva, operating) and Unit 3 (1405 MWe, KWU/Areva, construction). The PSR2-A1 attends the Standards 1.26-Safety in Operation of Nuclear Power Plants, Brazilian Nuclear Regulatory Commission (CNEN), and IAEA.SSG.25-Periodic Safety Review of Nuclear Power Plants. Within 18 months after each 10 years operation, the operating organization shall perform a plant safety review, to investigate the evolution consequences of safety code and standards, regarding: Plant design; structure, systems and components behavior; equipment qualification; plant ageing management; deterministic and probabilistic safety analysis; risk analysis; safety performance; operating experience; organization and administration; procedures; human factors; emergency planning; radiation protection and environmental radiological impacts. The Review included 6 Areas and 14 Safety Parameters, covered by 33 Evaluations.After document evaluations and discussions with plant staff, it was generated one General and 33 Specific Guide Procedures, 33 Specific and one Final Report, including: Description, Strengths, Deficiencies, Areas for Improvement and Conclusions. An Action Plan was prepared by Electronuclear for the recommendations. It was concluded that the Unit was operated within safety standards and will attend its designed operational lifetime, including possible life extensions. The Final Report was submitted to CNEN, as one requisite for renewal of the Unit Permanent Operation License. (author)

Second periodic safety review of Angra Nuclear Power Station, unit 1

International Nuclear Information System (INIS)

Martins, Carlos F.O.; Crepaldi, Roberto; Freire, Enio M.; Campello, Sergio A.

2015-01-01

This paper describes the second Periodic Safety Review (PSR2-A1) of Angra Nuclear Power Station, Unit 1, prepared by Eletrobras Eletronuclear S.A. and Tecnatom do Brasil Engenharia e Servicos Ltda., during Jul.2013-Aug.2014, covering the period of 2004-2013. The site, in Angra dos Reis-RJ, Brazil, comprises: Unit 1, (640 MWe, Westinghouse PWR, operating), Unit 2 (1300 MWe, KWU/Areva, operating) and Unit 3 (1405 MWe, KWU/Areva, construction). The PSR2-A1 attends the Standards 1.26-Safety in Operation of Nuclear Power Plants, Brazilian Nuclear Regulatory Commission (CNEN), and IAEA.SSG.25-Periodic Safety Review of Nuclear Power Plants. Within 18 months after each 10 years operation, the operating organization shall perform a plant safety review, to investigate the evolution consequences of safety code and standards, regarding: Plant design; structure, systems and components behavior; equipment qualification; plant ageing management; deterministic and probabilistic safety analysis; risk analysis; safety performance; operating experience; organization and administration; procedures; human factors; emergency planning; radiation protection and environmental radiological impacts. The Review included 6 Areas and 14 Safety Parameters, covered by 33 Evaluations.After document evaluations and discussions with plant staff, it was generated one General and 33 Specific Guide Procedures, 33 Specific and one Final Report, including: Description, Strengths, Deficiencies, Areas for Improvement and Conclusions. An Action Plan was prepared by Electronuclear for the recommendations. It was concluded that the Unit was operated within safety standards and will attend its designed operational lifetime, including possible life extensions. The Final Report was submitted to CNEN, as one requisite for renewal of the Unit Permanent Operation License. (author)

Evaluation of Perry Nuclear Power Plant Unit 1 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-11-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Perry Nuclear Power Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Perry T/S. Several discrepancies were identified and subsequently resolved through telephone conversations with the staff reviewer and the utility representative. Pending completion of the resolutions noted in Parts 3 and 4 of this report, the Perry Nuclear Power Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Assessment on Event Classification of One Steam Generator Tube Rupture in EU-APR

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hwan; Kim, Yong Soo [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Fukushima Daiichi nuclear power plant accident showed the vulnerability of coping strategy to beyond design natural disaster such as beyond design earthquake and tsunami. In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. Recently, the stress tests for WOLSONG Unit 1 and KORI Unit 1 have been performed and their assessment results have been reviewed by Korean regulatory body. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings. The several additional items for safety improvement has been drawn to enhance the coping capability for loss of safety functions under beyond design natural disaster in addition to post Fukushima action items.

Assessment on Event Classification of One Steam Generator Tube Rupture in EU-APR

International Nuclear Information System (INIS)

Kim, Ji Hwan; Kim, Yong Soo

2016-01-01

The Fukushima Daiichi nuclear power plant accident showed the vulnerability of coping strategy to beyond design natural disaster such as beyond design earthquake and tsunami. In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. Recently, the stress tests for WOLSONG Unit 1 and KORI Unit 1 have been performed and their assessment results have been reviewed by Korean regulatory body. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings. The several additional items for safety improvement has been drawn to enhance the coping capability for loss of safety functions under beyond design natural disaster in addition to post Fukushima action items

75 FR 8757 - Nebraska Public Power District, Cooper Nuclear Station, Unit 1; Notice of Availability of the...

Science.gov (United States)

2010-02-25

..., Cooper Nuclear Station, Unit 1; Notice of Availability of the Draft Supplement 41 to the Generic... Renewal of Cooper Nuclear Station, Unit 1 Notice is hereby given that the U.S. Nuclear Regulatory... operating license DPR-46 for an additional 20 years of operation for Cooper Nuclear Station, Unit 1 (CNS-1...

An integrated framework for effective reduction of occupational radiation exposure in a nuclear power plant

International Nuclear Information System (INIS)

Joo, Hyun Moon; Hak, Soo Kim; Young, Ho Cho; Chang, Sun Kang

1998-01-01

For effective reduction of occupational radiation exposure in a nuclear power plant, it is necessary to identify repetitive high radiation jobs during maintenance and refueling operation and comprehensively assess them. An integrated framework for effective reduction of occupational radiation exposure is proposed in this study. The framework consists of three parts; data collection, statistical analysis, and ALARA findings. A PC-based database program, INSTORE, is used for data collection and reduction, and the Rank Sum Method is used in identifying high radiation jobs. As a case study, the data accumulated in Kori Units 3 and 4 have been analyzed. The results of this study show that the radiation job classifications of SG related work have much effect on annual ORE collective dose in Kori Units 3 and 4. As an example of ALARA findings, hence, the improvements for the radiation job classifications of SG related work are summarized

Svakodnevna militarizacija života: etički aspekti korištenja djece u ratu

OpenAIRE

Rupčić, Darija

2017-01-01

Namjera je rada ukazati na problematiku izmijenjene naravi ratovanja u posljednjih nekoliko desetljeća 20. i 21. stoljeća, s osobitim naglaskom na problem korištenja sve većeg broja djece ratnika. Osnovna je teza rada ta da je praksa korištenja i regrutiranja djece u oružanim sukobima diljem svijeta najmanje prepoznata i najviše zanemarena forma zlostavljanja djece u suvremenom društvu te da je ona manje stvar kulture i nepostojanja stave društva prema vrijednostima djeteta, a više stvar prag...

Conformance to Regulatory Guide 1.97, Arkansas Nuclear One, Unit No. 1

International Nuclear Information System (INIS)

Stoffel, J.W.

1985-08-01

This EG and G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit No. 1 of Arkansas Nuclear One and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified

Study on the improvement of nuclear fuel cladding reliability

International Nuclear Information System (INIS)

Rheem, Karp Soon; Han, Jung Ho; Jeong, Yong Hwan; Lee, Deok Hyun

1987-12-01

In order to improve the nuclear fuel cladding reliability for high burn-up fuels, the corrosion resistance of laser beam surface treated and β-quenched zircaloys and the mechanical characteristics including fatigue, burst, and out-of-pile PCMI characteristics of heat treated zircaloys were investigated. In addition, the inadiation characteristics of Ko-Ri reactor fuel claddings was examined. It was found that the wasteside corrosion resistance of commercial zircaloys was improved remarkably by laser beam surface treatment. The out-of-pile transient cladding failures were investigated in terms of hoop stress versus time-to-failures by means of mandrel loading units at 25 deg C and 325 deg C. Fatigue characteristics of the β-quenched and as-received zircaloy cladding were investigated by using an internal oil pressurization method which can simulate the load-following operation cycle. The results were in good agreement with the existing data obtained by conventional methods for commercial zircaloys. Burst tests were performed with commercial and the β-quenched zircaloys in high pressure argon gas atmosphere as a function of burst temperature. The burst stress decreased linearly in the α phase region up to 600 deg C and hereafter the decrement of the burst stress decreased gradually with temperature in the β-phase region. For the first time, the burst characteristic of the irradiated zircaloy-4 cladding tubes released from Ko-Ri nuclear power unit 1 was investigated, and attempts were made to trace the cause of cladding failures by examining the failed structure and fret marks by debris. (Author)

Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

International Nuclear Information System (INIS)

Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

2011-01-01

Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

Design and implementation of an advanced protection system for the Shin-Kori 3 and 4 nuclear power plant

International Nuclear Information System (INIS)

Kim, Yonghak; Choi, Woongseock; Kwon, Jongsoo; Wilkosz, Stephen J.; Ridolfo, Charles F.; Yanosy, Paul L.

2008-01-01

The Nuclear Power Industry is currently embracing modern digital technology for upgrades to existing Instrumentation and Control (I and C) infrastructures as well as for incorporation into the next generation of new plants which will be coming 'on-line' during the next decade. This technology is being fully exploited for the next generation of advanced plant protection systems which will be initially deployed on the Shin-Kori 3 and 4 Nuclear Power Plant in the Republic of Korea. The system design for this plant protection system is being performed by the Korea Power Engineering Company (KOPEC) and builds upon the past generation of digital safety systems which were initially implemented at Ulchin 5 and 6. The advanced protection system is an evolution of this existing design and includes a number of improved operating attributes including: · Integration of Reactor Protection, Engineered Safety Features Actuation, and Qualified Indication and Alarm functions which were previously implemented by separate systems in the past. · Use of a 'soft control' interface which provides convenient accessibility to the safety systems from 'operator workstations' · Implementation of a Large Display Panel (LDP) which provides a consistent and constant representation of the overall plant state and of the plant safety status. The equipment for the advanced plant protection system is being provided by Westinghouse Electric Company (WEC) and utilizes the Westinghouse 'Common Q' Standardized qualified platform (where 'Q' denotes 'qualified'). The 'Common Q' platform is comprised of commercially dedicated Programmable Logic Controllers (PLC's), color-graphic Flat Panel Displays (FPD's) with integral touch screens, and high speed data communication links. It is a mature product that is in wide use for a number of safety-related applications. Among its key attributes are: · High overall system availability, which is achieved via use of a multiple channel configuration that is tolerant

Metallurgical characteristics and fracture mechanical properties of unirradiated Kori-1 RPV weld: Linde 80, WF-233

International Nuclear Information System (INIS)

Hong, Jun Hwa; Lee, B. S.; Oh, Y. J.; Chi, S. H.; Kim, J. H.; Park, D. G.; Yoon, J. H.; Oh, J. M.

2000-07-01

The fracture toughness transition properties of the low upper shelf weld, Linde 80 WF-233, of Kori-1 RPV were evaluated by the master curve method, which is designated by ASTM E 1921, 'Standard test method for determination of reference temperature, T o , for ferritic steels in the transition range'. The reference temperature, T o =-83 deg C, was determined by PCVN specimens at -90 deg C. This value is similar to that of other high copper welds. The initial RT NDT was conservatively estimated as -26 deg F from the current fracture toughness results. From the studies on the chemistry and microstructure, the fracture mechanical properties of WF-233 weld is convincingly not worse than WF-70 and 72W welds

75 FR 80547 - Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit No. 1; Exemption

Science.gov (United States)

2010-12-22

..., Shearon Harris Nuclear Power Plant, Unit No. 1; Exemption 1.0 Background Carolina Power & Light Company... operation of the Shearon Harris Nuclear Power Plant (HNP), Unit 1. The license provides, among other things... request to generically extend the rule's compliance date for all operating nuclear power plants, but noted...

Calculation of risk-based detection limits for radionuclides in the liquid effluents from Korean nuclear power plants

International Nuclear Information System (INIS)

Cheong, Jae Hak

2017-01-01

In order to review if present detection limits of radionuclides in liquid effluent from nuclear power plants are effective enough to warrant compliance with regulatory discharge limits, a risk-based approach is developed to derive a new detection limit for each radionuclide based on radiological criteria. Equations and adjustment factors are also proposed to discriminate the validity of the detection limits for multiple radionuclides in the liquid effluent with or without consideration of the nuclide composition. From case studies to three nuclear power plants in Korea with actual operation data from 2006 to 2015, the present detection limits have turned out to be effective for Hanul Unit 1 but may not be sensitive enough for Kori Unit 1 (8 out of 14 radionuclides) and Wolsong Unit 1 (9 out of 42 radionuclides). However, it is shown that the present detection limits for the latter two nuclear power plants can be justified, if credit is given to the radionuclide composition. Otherwise, consideration should be given to adjustment of the present detection limits. The risk-based approach of this study can be used to determine the validity of established detection limits of a specific nuclear power plant. (author)

Evaluation of applicability of alternative source terms to operating nuclear power plants in Korea

International Nuclear Information System (INIS)

Lim, S. N.; Park, Y. S.; Nam, K. M.; Song, D. B.; Bae, Y. J.; Lee, Y. J.; Jung, C. Y.

2002-01-01

In 1995 and 2000, NRC issued NUREG-1465 and Regulatory Guide 1.183 with respect to Alternative Source Terms(AST) replacing the existing source terms of TID-14844 and Regulatory Guide 1.4, 1.25, and 1.77 for radiological Design Basis Accidents(DBA) analysis. In 1990, ICRP published ICRP Pub. 60 which represents new recommendations on dose criteria and concepts. In Korea, alternative source terms were used for evaluation of effective doses for design basis accidents of Advanced Power Reactor(APR1400) using the computer program developed by an overseas company. Recently, DBADOSE, new computer program for DBA analysis incorporating AST and effective dose concept was developed by KHNP and KOPEC, and reanalysis applying AST to operating nuclear power plants, Kori units 3 and 4 in Korea using DBADOSE has been performed. As the results of this analysis, it was concluded that some conservative variables or operation procedures of operating plants could be mitigated or simplified by virtue of increased safety margin and consequently, economical and operational benefits ensue. In this paper, methodologies and results of Kori 3 and 4 DBA reanalysis and sensitivity analysis for mitigation of main design variables are introduced

The licensing practice on nuclear power plants in Korea

International Nuclear Information System (INIS)

Moon, S. P.

1994-01-01

The evolution of Korean regulatory system has tightly coupled with development ot Korean nuclear power program. The nuclear power plant licensing has become a major regulatory function of the government when the construction of the Kori NPP Unit 1 started in early 1970s. During this period, domestic laws and regulations applicable to the licensing of NPP were not yet fully developed. Therefore the vendor countries' laws and regulations were applied as mandatory requirement. Beginning in the early 19808, component approach was used and contracts were awarded separately for major components of the plants, thus enabling more domestic industries to participate in the projects. The two-step licensing system was incorporated into the law. In the third phase from 1987, major efforts have been concentrated on the maximum participation of local industries. The overriding priority for selecting suppliers was the condition of higher nuclear technology transfer to Korea. The Korea Institute of Nuclear Safety (KINS) was established in 1990 as an independent regulatory expert organization

75 FR 6223 - PSEG Nuclear LLC; Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1...

Science.gov (United States)

2010-02-08

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-272, 50-311 and 50-354; NRC-2010-0043] PSEG Nuclear LLC; Hope Creek Generating Station and Salem Nuclear Generating Station, Unit Nos. 1 and 2...-70, and DPR-75, issued to PSEG Nuclear LLC (PSEG, the licensee), for operation of the Hope Creek...

76 FR 19148 - PSEG Nuclear, LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1...

Science.gov (United States)

2011-04-06

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-272, 50-311, 50-354; NRC-2009-0390 and NRC-2009-0391] PSEG Nuclear, LLC, Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2..., DPR-70, and DPR-75 for an additional 20 years of operation for the Hope Creek Generating Station (HCGS...

The licensing practice on nuclear power plants in Korea

Energy Technology Data Exchange (ETDEWEB)

Moon, S P

1994-12-31

The evolution of Korean regulatory system has tightly coupled with development ot Korean nuclear power program. The nuclear power plant licensing has become a major regulatory function of the government when the construction of the Kori NPP Unit 1 started in early 1970s. During this period, domestic laws and regulations applicable to the licensing of NPP were not yet fully developed. Therefore the vendor countries` laws and regulations were applied as mandatory requirement. Beginning in the early 19808, component approach was used and contracts were awarded separately for major components of the plants, thus enabling more domestic industries to participate in the projects. The two-step licensing system was incorporated into the law. In the third phase from 1987, major efforts have been concentrated on the maximum participation of local industries. The overriding priority for selecting suppliers was the condition of higher nuclear technology transfer to Korea. The Korea Institute of Nuclear Safety (KINS) was established in 1990 as an independent regulatory expert organization. 1 tab., 4 figs.

Development of a Pilot Program for Human Factors Management in Operating Nuclear Power plants

International Nuclear Information System (INIS)

Lee, Jung-Woon; Lee, Yong-Hee; Jang, Tong-Il; Kim, Dae-Ho

2007-01-01

The human factors of operating NPPs have been reviewed as a part of Periodic Safety Reviews (PSRs). This human factors PSR covers a wide range of human factors including control room man-machine interfaces (MMIs), procedures, working conditions, qualification, training, information requirements and workload. Korea Atomic Energy Research Institute (KAERI) has performed human factors PSRs from the first PSR for Kori 1. It was determined in 2005 that for a Continuous Operation of the Korean NPPs an enhanced PSR should be performed and issues raised from the PSRs should be resolved. From the results of the PSR for Kori 1, several safety enhancement issues related to human factors were raised. KAERI is working on a resolution of some of the human factors issues for the Korea Hydro and Nuclear Power Co. (KHNP). As a part of the resolution, we are developing a human factors management program (HFMP) for Kori 1. This paper introduces the status of our development of HFMP

FIND: Douglas Point Nuclear Generating Station, Units 1 and 2

International Nuclear Information System (INIS)

Moore, M.M.

1975-12-01

This index is presented as a guide to microfiche items 1 through 136 in Docket 50448, which was assigned to Potomac Electric Power Company's Application for Licenses to construct and operate Douglas Point Nuclear Generating Station, Units 1 and 2. Information received from August, 1973 through July, 1975 is included

Nuclear Power Plant Lifetime Management Study (I)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

Nuclear Power Plant Lifetime Management Study (I)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-12-31

As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

Summary revaluation of energetic start-up of the unit 1 of nuclear power plant Mochovce

International Nuclear Information System (INIS)

Sarvaic, I.; Miskolci, M.

1998-01-01

The document contents stage revaluation of energetic start-up of the unit 1 of nuclear power plant Mochovce. Test results of the stage of energetic start-up are summarized in the document, valuation of important systems and block devices as well as fulfilling the operation limits and conditions has been performed. On that base conclusions and recommendations for start-up the unit 2 and for commercial operation of the unit 1 are elaborated. The valuation has been elaborated by a scientific management for start-up nuclear power plant Mochovce of nuclear safety of nuclear power facilities. Scientific management for start-up of nuclear power plant Mochovce performed continuous valuation of individual power levels after ending of each individual level and it gave its valuation to energy power level with recommendations and conditions for further start-up process and operation. Scientific management finished its activity at the unit 1 of nuclear power plant Mochovce according to a statute of scientific management for start-up after successful completion of conclusive block run. Scientific management group was founded in February 1998 at nuclear power plant Mochovce. Its members are experts from Slovak, Czech, Russian and French organizations which are participating in power plant completion. Members are listed in a supplement No. 2

The Economic Value of Korean Nuclear Power Industry in the National Economy: An Input-Output Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, M. K.; Kim, S. S.; Lee, J. H.; Kim, S. H. [Nuclear Policy Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

In 1978, Korea introduced the first nuclear power plant, Kori-1 unit, in parallel with the nation's industrialization policy. Thereafter, Korea has carried out a very ambitious nuclear power program and sustained a strong commitment to nuclear power development. Thus, nuclear is a prime energy source which presently meets about 30 percent of Korea's power demands. Also, Korea won a contract for APR-1400 NPPs to the UAE in 2009 which led to Korea as a significant exporter in the world nuclear market. Recently, the new government of Korea has been launching 'Creative Economy', from this perspective, the quantitative contributions of nuclear sector to the national economic growth are required to be estimated. This paper is to estimate quantitatively the economic values created by nuclear power industry in the framework of national economy. The total economic values created by nuclear power industry are estimated to be 63.6 trillion won for the study period.

Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Abbott, L [ed.

1985-09-01

An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures.

Pressurized thermal shock evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant

International Nuclear Information System (INIS)

Abbott, L.

1985-09-01

An evaluation of the risk to the Calvert Cliffs Unit 1 nuclear power plant due to pressurized thermal shock (PTS) has been completed by Oak Ridge National Laboratory (ORNL) with the assistance of several other organizations. This evaluation was part of a Nuclear Regulatory Commission program designed to study the PTS risk to three nuclear plants, the other two plants being Oconee Unit 1 and H.B. Robinson Unit 2. The specific objectives of the program were to (1) provide a best estimate of the frequency of a through-the-wall crack in the pressure vessel at each of the three plants, together with the uncertainty in the estimated frequency and its sensitivity to the variables used in the evaluation; (2) determine the dominant overcooling sequences contributing to the estimated frequency and the associated failures in the plant systems or in operator actions; and (3) evaluate the effectiveness of potential corrective measures

75 FR 43571 - Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; Environmental Assessment And...

Science.gov (United States)

2010-07-26

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-413 and 50-414; NRC-2010-0260] Duke Energy Carolinas, LLC; Catawba Nuclear Station, Units 1 and 2; Environmental Assessment And Finding of No Significant... Energy Carolinas, LLC (the licensee), for operation of the Catawba Nuclear Station, Units 1 and 2...

Audit Technical of Kori Rubber Dam in the River of Keyang District of Ponorogo East Java Province

Science.gov (United States)

Murnianto, E.; Suprapto, M.; Ikhsan, C.

2018-03-01

The development of science and technology for the utilization and protection of rivers has embodied various types of river infrastructure. Without proper maintenance, rapid river sediments undergo physical degradation and function. Problems that occur in Kori Rubber Dam, among others, the damage to the body of the rubber dam that is made of rubber, so that the function of flower deflection is not optimal. This happens because of limited operational and maintenance activities (OM). A technical audit is a process of identifying problems, analyzing, and evaluating ones conducted independently, objectively and professionally on the basis of examination, to assess the truth, accuracy, credibility, and reliability of information about a job. In this case an assessment of the Kori Rubber Dam, which is basically a benchmarking activity. Assessment of rubber dam components includes the physical conditions and functions that affect the weir. This research is expected to know the performance of Kori rubber Dam as a recommendation material in the implementation of OM Rubber Dam activities.

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.

78 FR 49305 - Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2...

Science.gov (United States)

2013-08-13

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-445 and 50-446; NRC-2013-0182] Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2; Application for Amendment to Facility... Operating License Nos. NPF-87 and NPF-89 for the Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2...

76 FR 72007 - ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security...

Science.gov (United States)

2011-11-21

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-295 and 50-304; NRC-2011-0244] ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor...

76 FR 24064 - Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Notice...

Science.gov (United States)

2011-04-29

... Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Notice of Issuance of Renewed... Company (licensee), the operator of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (PVNGS... Plants: Supplement 43, Regarding Palo Verde Nuclear Generating Station,'' issued January 2011, discusses...

Status of Korean nuclear industry and Romania-Korea cooperation in the field of nuclear power

International Nuclear Information System (INIS)

Lee, Myung Key

2005-01-01

The Kyoto Protocol on climate change has urged the world to explore ways of cutting down the greenhouse emissions, and it also boosted a number of nuclear power projects that is so-called the renaissance of nuclear power. Nuclear power has proven to be the cleanest energy source and one of the cheapest types of energies, compared with other energy sources. Korea began developing its nuclear power projects from the early 1970's. Since the first nuclear power plant Kori Unit 1, started commercial operation in 1978, Korea has continuously promoted the development of nuclear power projects, and today it operates 20 nuclear power units (17,716 MW), including 4 units of CANDU plants. Korea ranked No. 6 in the world in terms of installed capacity of nuclear power plants, and 40% of its domestic electricity generation comes from nuclear power plants. The average plant capacity factor was 95.5% in 2005, which is about 16% than the world average of around 79%. All the Korean nuclear power projects are led and implemented by Korea Hydro and Nuclear Power Co. (KHNP) which is the sole state-owned nuclear power project company spun off from Korea Electric Power Corporation (KEPCO) in 2001 as part of the government's program for electric industry restructuring. The cooperation between Romania and Korea in the nuclear power field began in March 2001. At industrial level a technical agreement between the Romanian Company Nuclearelectrica S.A. (SNN) and KHNP was signed in July 2003 for cooperation in Cernavoda NPP projects. The joint development of the Cernavoda NPP unit 3 was one of the major topics. Heavy water produced by Romanian Heavy Water plant at Drobeta Turnu Severin was supplied to KHNP (16 tones in 2001 and another 16 tones in 2004). The feasibility study for units 3 and 4 is being performed in two phases under leadership of SNN in cooperation with KHNP, AECL, ANSALDO and Deloitte and Touche as a financial advisor in Phase 2. It is expected that the appropriate securities

75 FR 43572 - Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2; Environmental Assessment and...

Science.gov (United States)

2010-07-26

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-369 and 50-370; NRC-2010-0259] Duke Energy Carolinas, LLC, McGuire Nuclear Station, Units 1 and 2; Environmental Assessment and Finding of No Significant... Energy Carolinas, LLC (the licensee), for operation of the McGuire Nuclear Station, Units 1 and 2...

Introduction to Nuclear Power Plant Environment Supervisory Committee

International Nuclear Information System (INIS)

Lee, Byung Il

2008-01-01

In Korea, there are five nuclear power plant sites, located at Yongkwang, Kori, Shin Kori, Uljin, and Wolsong. Each administrative district has its own NESC (Nuclear plant Environment Public Supervisory Committee) which consists of a steering committee and a center. The purpose of NESC is to let the public survey and inspect nuclear plant environment and then improve a clarity and confidence in plant construction and operation by themselves. In order to understand the situation of NESC and explore ways toward a better role of NESC, in this paper we try to enumerate a few major facts related to a current status of NESC. As a summary, there must be a great role of NESC in the relationship between the residents, a nuclear industry company and the related government division. Furthermore, NESC would certainly do its role for all sides provided that more strong definition of NESC in law, more financial supporting and more philosophical speculation for the being of NESC

The relaxation of ESFAS/RPS surveillance test requirements

International Nuclear Information System (INIS)

Hah, Yung Joon; Koo, Jung Eui; Choi, Hae Yoon

1994-01-01

The surveillance test requirement of ESFAS/RPS is reviewed for 950 MWe class westinghouse reactor (YGN unit 1 and 2, Kori unit 3 and 4). The current requirements of frequent test and maintenance in the tech. spec. can lead to human errors, jeopardizing safety of the plant, and reduction in the availability of the plant. Meanwhile, the ESFAS designs do not provide for complete online testing capabilities for their protection systems. Therefore, ESFAS slave relays cannot be tested during plant operation as actuation of associated equipment could result in unwanted plant transient or trip conditions. In this study, westinghouse's PSA results, NRC recommendation and NRC approval status for specific U.S. nuclear power plant have been reviewed and evaluated. Since YGN 1 and 2 and Kori 3 and 4 are essentially the same plant as the operating westinghouse plant in the U.S., it is expected that YGN 1 and 2 and Kori 3 and 4 will be justified for having ESFAS/RPS surveillance test requirements relaxation program. Finally the extension of surveillance testing intervals and allowed outage times for test and maintenance will be verified by PSA program for YGN 1 and 2 and Kori 3 and 4. Various findings during the project can be used in the possible future revision of current technical specification with further refinements through the PSA program for YGN 1 and 2 and Kori 3 and 4. (Author) 10 refs., 8 figs., 24 tabs

The relaxation of ESFAS/RPS surveillance test requirements

Energy Technology Data Exchange (ETDEWEB)

Hah, Yung Joon; Koo, Jung Eui; Choi, Hae Yoon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-01-01

The surveillance test requirement of ESFAS/RPS is reviewed for 950 MWe class westinghouse reactor (YGN unit 1 and 2, Kori unit 3 and 4). The current requirements of frequent test and maintenance in the tech. spec. can lead to human errors, jeopardizing safety of the plant, and reduction in the availability of the plant. Meanwhile, the ESFAS designs do not provide for complete online testing capabilities for their protection systems. Therefore, ESFAS slave relays cannot be tested during plant operation as actuation of associated equipment could result in unwanted plant transient or trip conditions. In this study, westinghouse`s PSA results, NRC recommendation and NRC approval status for specific U.S. nuclear power plant have been reviewed and evaluated. Since YGN 1 and 2 and Kori 3 and 4 are essentially the same plant as the operating westinghouse plant in the U.S., it is expected that YGN 1 and 2 and Kori 3 and 4 will be justified for having ESFAS/RPS surveillance test requirements relaxation program. Finally the extension of surveillance testing intervals and allowed outage times for test and maintenance will be verified by PSA program for YGN 1 and 2 and Kori 3 and 4. Various findings during the project can be used in the possible future revision of current technical specification with further refinements through the PSA program for YGN 1 and 2 and Kori 3 and 4. (Author) 10 refs., 8 figs., 24 tabs.

Trace of nuclear energy with pictures

International Nuclear Information System (INIS)

1992-05-01

This book traces the history of development over nuclear energy with pictures, which contains preface, development history of the world, development history of Korea, nuclear power plant in Kori, nuclear power plant in Wolseong, nuclear power plant in Yeonggwang, nuclear power plant in Uljin, nuclear fuel, using of radiation and radioactive isotope, development of nuclear energy in the world and a Chronological table of nuclear energy. This book is written to record the development history of Korea through pictures of the nuclear power plants in Korea.

Station black out of Fukushima Daiichi Nuclear Power Station Unit 1 was not caused by tsunamis

International Nuclear Information System (INIS)

Ito, Yoshinori

2013-01-01

Station black out (SBO) of Fukushima Daiichi Nuclear Power Station Unit 1 would be concluded to be caused before 15:37 on March 11, 2011 because losses of emergency ac power A system was in 15:36 and ac losses of B system in 15:37 according to the data published by Tokyo Electric Power Co. (TEPCO) in May 10, 2013. Tsunami attacked the site of Fukushima Daiichi Nuclear Power Station passed through the position of wave amplitude meter installed at 1.5 km off the coast after 15:35 and it was also recognized tsunami arrived at the coast of Unit 4 sea side area around in 15:37 judging from a series of photographs taken from the south side of the site and general knowledge of wave propagation. From a series of photographs and witness testimony, tsunami didn't attack Fukushima Daiichi Nuclear Power Station uniformly and tsunami's arrival time at the site of Unit 1 would be far later than arrival time at the coast of Unit 4 sea side area, which suggested it would be around in 15:39. TEPCO insisted tsunami passed through 1.5 km off the coast around in 15:33 and clock of wave amplitude meter was incorrect, which might be wrong. Thus SBO of Fukushima Daiichi Nuclear Power Station Unit 1 occurred before tsunami's arrival at the site of Unit 1 and was not caused by tsunami. (T. Tanaka)

Start up and commercial operation of Laguna Verde nuclear power plant. Unit 1

International Nuclear Information System (INIS)

Torres Ramirez, J.F.

1991-01-01

Prior to start up of Laguna Verde nuclear power plant preoperational tests and start tests were performed and they are described in its more eminent aspects. In relation to commercial operation of nuclear station a series of indicator were set to which allow the measurement of performance in unit 1, in areas of plant efficiency and personal safety. Antecedents. Laguna Verde station is located in Alto Lucero municipality in Veracruz state, 70 kilometers north-northeast from port of Veracruz and a 290 kilometers east-northeast from Mexico city. The station consist of two units manufactured by General Electric, with a nuclear system of vapor supply also called boiling water (BWR/5), and with a system turbine-generator manufactured by Mitsubishi. Each unit has a nominal power of 1931 MWt and a level design power of 675 Mwe and a net power of 654 Electric Megawatts

ATWS analysis for Browns Ferry Nuclear Plant Unit 1

International Nuclear Information System (INIS)

Dallman, R.J.; Jouse, W.C.

1985-01-01

Analyses of postulated Anticipated Transients Without Scram (ATWS) were performed at the Idaho National Engineering Laboratory (INEL). The Browns Ferry Nuclear Plant Unit 1 (BFNP1) was selected as the subject of this work because of the cooperation of the Tennessee Valley Authority (TVA). The work is part of the Severe Accident Sequence Analysis (SASA) Program of the US Nuclear Regulatory Commission (NRC). A Main Steamline Isolation Valve (MSIV) closure served as the transient initiator for these analyses, which proceeded a complete failure to scram. Results from the analyses indicate that operator mitigative actions are required to prevent overpressurization of the primary containment. Uncertainties remain concerning the effectiveness of key mitigative actions. The effectiveness of level control as a power reduction procedure is limited. Power level resulting from level control only reduce the Pressure Suppression Pool (PSP) heatup rate from 6 to 4 0 F/min

I and C success in Korea past and present projects and a view to the future

International Nuclear Information System (INIS)

Ridolfo, Charles F.

2009-01-01

Westinghouse Electric Company, LLC (WEC) has been providing advanced Instrumentation and Control (I and C) technology for the C E System 80, Korean Standard Nuclear Power Plant (KSNP), since the early 1990's. I and C equipment and technical support has been provided for a progression of these plants including Yonggwang Units 3,4,5,6, Ulchin Units 3,4,5,6, Shin Kori Units 1 and 2, Shin Wolsong Units 1 and 2, and most recently for Shin Kori Units 3 and 4. These I and C programs have been highly successful and have helped to contribute to the outstanding plant availability for the commissioned units at Yonggwang 3,4,5, and 6 and Ulchin 3,4,5, and 6; which are currently providing power on the grid. WEC has also recently delivered I and C equipment for Shin Kori Units 1 and 2 and Shin Wolsong Units 1 and 2 and is supporting the installation and startup of the equipment. In addition, WEC is in the initial design stages for providing an integrated protection system that will be deployed at Shin Kori Units 3 and 4 which are the first units to implement the APR1400; an advanced reactor of indigenous Korean design, which is based on the previous generation Optimized Power Reactor OPR1000. The success of the I and C programs has been the result of careful consideration of the appropriate technology to employ, applying comprehensive quality assurance measures, providing appropriate technical support and consultation services, implementation of a program of continued I and C logistic support, maintaining a professional and experienced I and C work force, and maintaining a strong and mutually supportive partnership with the Korea Power Engineering Company (KOPEC) and with Korea Hydro and Nuclear Power (KHNP). WEC takes great pride in its on going I and C partnership with Korea, which has included not only providing equipment and technical support services, but also completion of a comprehensive Technology Transfer program. This program was implemented to promote local 'elf

Aging management strategy for reactor internals of Korean nuclear power plants

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

2010-01-01

This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

Effect of Containment Spray Additives on the Chemical Effect after a Loss of Coolant Accident in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Chan; Park, Jong Woon; Lee, Guen Sung [KOREA HYDRO and NUCLEAR POWER Co., Daejeon (Korea, Republic of)

2007-10-15

As a part of USNRC GSI-191, evaluation of Kori Unit 1 ECCS recirculation sump performance has been carried out in 2006. The work is derived from the result of first PSR(Periodic Safety Review) of Kori Unit1. In this work, we have considered the replacement of spray additive in containment building to solve issues of GSI-191 and GL2004-02. We estimated the chemical effect of changing NaOH into TSP(Trisodium Phosphate) based on SRP(Standard Review Plan) 6.5.2. Rev.02. WCAP-16530 methodology is used to compare chemical effects of spray additive(or buffering agents). In the other side, chemical thermodynamic simulation can be utilized. Herein, the results using WCAP-16530 methodology and chemical simulation are presented.

Planning for decommissioning of Ignalina Nuclear Power Plant Unit-1

International Nuclear Information System (INIS)

Poskas, P.; Poskas, R.; Zujus, R.

2002-01-01

In accordance to Ignalina NPP Unit 1 Closure Law, the Government of Lithuania approved the Ignalina NPP Unit 1 Decommissioning Program until 2005. For enforcement of this program, the plan of measures for implementation of the program was prepared and approved by the Minister of Economy. The plan consists of two parts, namely technical- environmental and social-economic. Technical-environmental measures are mostly oriented to the safe management of spent nuclear fuel and operational radioactive waste stored at the plant and preparation of licensing documents for Unit 1 decommissioning. Social-economic measures are oriented to mitigate the negative social and economic impact on Lithuania, inhabitants of the region, and, particularly, on the staff of Ignalina NPP by means of creating favorable conditions for a balanced social and economic development of the region. In this paper analysis of planned radioactive waste management technologies, licensing documents for decommissioning, other technical-environmental and also social-economic measures is presented. Specific conditions in Lithuania important for defining the decommissioning strategy are highlighted. (author)

Decommissioning of units 1 - 4 at Kozloduy nuclear power plant in Bulgaria

International Nuclear Information System (INIS)

Dishkova, Denitsa

2014-01-01

Nuclear safety and security are absolute priorities for the European Union countries and this applies not only to nuclear power plants in operation but also to decommissioning. In terms of my technical background and my working experience in the field of licensing and environmental impact assessment during the decommissioning of Units 1 to 4 at Kozloduy Nuclear Power Plant (KNPP) in Bulgaria, I decided to present the strategy for decommissioning of Units 1 to 4 at KNPP which was selected and followed to achieve safe and effective decommissioning process. The selected strategy in each case must meet the legislative framework, to ensure safe management of spent fuel and radioactive waste, to provide adequate funding and to lead to positive socio-economic impact. The activities during the decommissioning generate large volume of waste. In order to minimize their costs and environmental impact it should be given a serious consideration to the choice, the development and the implementation of the most adequate process for treatment and the most appropriate measurement techniques. The licensing process of the decommissioning activities is extremely important and need to cope with all safety concerns and ensure optimal waste management. (authors)

Current status and future prospects of Korean standardized nuclear power plant design

International Nuclear Information System (INIS)

Rieh, C.-H.; Park, S.-K.; Lee, B.-R.

1992-01-01

The authors reviewed a brief history of Korean nuclear industry since the first Kori-1 plant operation in 1978 with special emphasis on the NSSS and BOP design and engineering, and the design approaches for nuclear power plants in the future. Continued effort to enhance plant economy and operational safety has been made by increasing plant size, and improving safety features, systems and component reliability in various design aspects. Korean nuclear industry is now trying to be one of the major contributors to the world nuclear field in sharing nuclear technology gained from past experience and developed through internation technical cooperation programs

Strategy of nuclear power in Korea, non-nuclear-weapon state and peaceful use of nuclear power

International Nuclear Information System (INIS)

Nagasaki, Takao

2005-01-01

The nuclear power plant started at Kori in Korea in April, 1978. Korea has carried out development of nuclear power as a national policy. The present capacity of nuclear power plants takes the sixes place in the world. It supplies 42% total power generation. The present state of nuclear power plant, nuclear fuel cycle facility, strategy of domestic production of nuclear power generation, development of next generation reactor and SMART, strategy of export in corporation with industry, government and research organization, export of nuclear power generation in Japan, nuclear power improvement project with Japan, Korea and Asia, development of nuclear power system with nuclear diffusion resistance, Hybrid Power Extraction Reactor System, radioactive waste management and construction of joint management and treatment system of spent fuel in Asia are stated. (S.Y.)

Preliminary study on the analysis of alpha emitters at working places in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Hanna; Kim, Jeong In; Lee, Byoungil [Korea Hydro and Nuclear Power Co., Seoul (Korea, Republic of); Yoon, Suk Won [Korea Institute of Radiological and Medical Sciences National Radiation Emergency Medical Center, Seoul (Korea, Republic of)

2012-10-15

Over sea nuclear power plants have been reported cases of internal contamination by alpha nuclides. In many cases, stations encountered significant alpha contamination when aged/legacy equipment was disturbed or handled. Under normal operating conditions, transuranic radionuclides are contained within the fuel rods and therefore are not a contributor to radioactive contamination within a facility. However, transuranic radionuclides result from the presence of tramp-uranium contamination on the exterior of fuel elements. Fuel failures may develop during operating cycles due to a variety of causes, ranging from manufacturing defects to mechanical or abrasive damage. In case of domestic nuclear power plants, the pressure tube replacements in Wolsong Unit 1 and steam generator replacements in Kori Unit 1 were done. Due to deterioration of equipment in accordance with the long-term operation, the domestic nuclear power plants are expected to improve the facilities and the probability of internal exposure from alpha emitters is increasing. The domestic nuclear power plants are only keeping alpha radionuclides of the effluent from the exterior under constant surveillance. The representative areas of CV are just carried out continuous alpha monitoring in during a unit outage. So far, there is no other case with alpha nuclides analysis. As the domestic nuclear power plants are expected to improve the facilities, it is the time to take proactive measures to deal with internal contamination by alpha emitting radioactive elements. In this paper, the possible risk of internal exposure is based on preliminary experiments on the analysis of alpha emitting radioactive elements at working places in nuclear power plants.

75 FR 16201 - FPL Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 1 and 2; Exemption

Science.gov (United States)

2010-03-31

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-266 and 50-301; NRC-2010-0123] FPL Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 1 and 2; Exemption 1.0 Background FPL Energy Point Beach.... Borchardt (NRC) to M. S. Fertel (Nuclear Energy Institute) dated June 4, 2009. The licensee's request for an...

Millstone Unit 1 plant vulnerabilities during postulated severe nuclear accidents

International Nuclear Information System (INIS)

Khalil, Y.F.

1993-01-01

Generic Letter 88-20, Supplement No. 1 (Ref. 1), issued by the Nuclear Regulatory Commission (NRC) requested all licensees holding operating licenses and construction permits for nuclear power reactor facilities to perform Individual Plant Examinations (IPE) of their plant(s) for severe accident vulnerabilities and to submit the results to the Commission. This paper summarizes the major Front-End (Level-1 PRA) and Back-End (Level-2 PRA) insights gained from the Millstone Unit 1 (MP-1) IPE study. No major plant vulnerabilities have been identified from a Front-End perspective. The Back-End analysis, however, has identified two potential containment vulnerabilities during postulated events that progress beyond the Design Basis Accidents (DBAs), namely, (1) MP-1 is dominated by early source term releases that would occur within a six-hour time frame from time of accident initiation, or reactor trip, and (2) MP-1 containment is somewhat vulnerable to leak-type failure through the drywell head. As a result of the second finding, a recommendation currently under evaluation, has been made to increase the drywell head bolt's preload from 54 Kips to resist the containment design pressure value (62 psig)

78 FR 14361 - In the Matter of Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Units 1 and...

Science.gov (United States)

2013-03-05

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0310; Docket Nos. 50-445 and 50-446; License Nos. NPF-87 and NPF-89] In the Matter of Luminant Generation Company LLC, Comanche Peak Nuclear Power Plant, Units... Nuclear Power Plant, Units 1 and 2 (CPNPP), and its Independent Spent Fuel Storage Installation Facility...

75 FR 8149 - Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

Science.gov (United States)

2010-02-23

...] Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3... NPF-74, issued to the Arizona Public Service Company (APS, or the licensee), for operation of the Palo Verde Nuclear Generating Station (PVNGS, the facility), Units 1, 2, and 3, respectively, located in...

Integrated-plant-safety assessment Systematic Evaluation program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245

International Nuclear Information System (INIS)

1982-11-01

The Systematic Evaluation Program was initiated in February 1977 to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

Interim reliability evaluation program: analysis of the Arkansas Nuclear One. Unit 1 Nuclear Power Plant

International Nuclear Information System (INIS)

Kolb, G.J.; Kunsman, D.M.; Bell, B.J.

1982-06-01

This report represents the results of the analysis of Arkansas Nuclear One (ANO) Unit 1 nuclear power plant which was performed as part of the Interim Reliability Evaluation Program (IREP). The IREP has several objectives, two of which are achieved by the analysis presented in this report. These objectives are: (1) the identification, in a preliminary way, of those accident sequences which are expected to dominate the public health and safety risks; and (2) the development of state-of-the-art plant system models which can be used as a foundation for subsequent, more intensive applications of probabilistic risk assessment. The primary methodological tools used in the analysis were event trees and fault trees. These tools were used to study core melt accidents initiated by loss of coolant accidents (LOCAs) of six different break size ranges and eight different types of transients

75 FR 18572 - Supplemental Environmental Impact Statement for Sequoyah Nuclear Plant Units 1 and 2 License...

Science.gov (United States)

2010-04-12

... TENNESSEE VALLEY AUTHORITY Supplemental Environmental Impact Statement for Sequoyah Nuclear Plant... National Environmental Policy Act. TVA will prepare a supplemental environmental impact statement (SEIS) to update information in the 1974 Final Environmental Statement for Sequoyah Nuclear Plant Units 1 and 2...

Nuclear design report for Ulchin nuclear power plant unit 1, cycle 7

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Rae; Park, Yong soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-04-01

This report presents nuclear design calculations for Cycle 7 of Ulchin Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 56 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s 36 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 7 amounts to 355 EFPD corresponding to a cycle burnup of 14280 MWD/MTU. (Author) 8 refs., 55 figs., 21 tabs.

Nuclear design report for Yonggwang nuclear power plant unit 1 cycle 9

Energy Technology Data Exchange (ETDEWEB)

Cho, Young chul; Kim, Jae Hak; Song, Jae Woong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-03-01

This report presents nuclear design calculations for Cycle 6 of Yonggwng Unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 76 KOFA`s enriched by nominally 4.00 w/o U{sub 235}. Among the KOFA`s, 60 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 9 amounts to 434 EFPD corresponding to a cycle burnup of 17470 MWD/MTU. (Author) 8 refs., 55 figs., 19 tabs.

Nuclear design report for Ulchin nuclear power plant unit 1, cycle 6

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Kyun; Kim, Yong Rae; Park, Yong Soo; Cho, Byeong Ho; Lee, Sang Keun; Ahn, Dawk Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-12-01

This report presents nuclear design calculations for cycle 6 of Ulchin unit 1. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 64 KOFA`s enriched by nominally 3.70 w/o U{sub 235}. Among the KOFA`s, 32 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 6 amounts to 369 EFPD corresponding to a cycle burnup of 14850 MWD/MTU. (Author) 8 refs., 55 figs., 17 tabs.

Technology development on the assessment of structural integrity of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Seo, Jeong Moon; Choun, Y. S.; Choi, I. K. and others

1999-04-01

Nuclear power plants in Korea show drop off in their performance and safety margin as the age of plants increase. The reevaluation of Kori-1 Unit on its performance and safety for life extension is expected in the near future. However, technologies and information related are insufficient to quantitatively estimate them. The final goal of this study is to develop the basic testing and evaluation techniques related with structural integrity of important nuclear equipment and structures. A part of the study includes development of equipment qualification technique. To ensure the structural integrity of structures, systems, and equipment in nuclear power plants, the following 5 research tasks were performed in the first year. - Analysis of dynamic characteristics of reactor internals - Analysis of engineering characteristics of instrumental earthquakes recorded in Korea - Analysis of ultimate pressure capacity and failure mode of containments building - Development of advanced NDE techniques using ultrasonic resonance scattering - Development of equipment qualification technique against vibration aging. These technologies developed in this study can be used to ensure the structural safety of operational nuclear power plants, and for the long-term life management. (author)

Nuclear power in the United States

International Nuclear Information System (INIS)

Johnston, J.B.

1985-01-01

All over the world except in the United States, nuclear energy is a low cost, secure, environmentally acceptable form of energy. In the United States, civilian nuclear power is dead. 112 nuclear power plants have been abandoned or cancelled in the last decade, and there has been no new order for nuclear plants since 1978. It will be fortunate to have 125 operating nuclear plants in the United States in the year 2000. There are almost 90 completed nuclear power plants and about 45 under construction in the United States, but several of those under construction will eventually be abandoned. About 20 % of the electricity in the United States will be generated by nuclear plants in 2000 as compared with 13 % supplied in the last year. Under the present regulatory and institutional arrangement, American electric utilities would not consider to order a new nuclear power plant. Post-TMI nuclear plants became very expensive, and there is also ideological opposition to nuclear power. Coal-firing plants are also in the similar situation. The uncertainty about electric power demand, the cost of money, the inflation of construction cost and regulation caused the situation. (Kako, I.)

Nuclear power plant V-1

International Nuclear Information System (INIS)

1998-01-01

The nuclear power plant Bohunice V -1 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - 24 April 1972; first controlled reactor power - 27 November 1978, 15 March 1980; connection to the grid - 17 December 1978, 26 March 1980; commercial operation - 1 April 1980, 7 January 1981. This leaflet contains: NPP V-1 construction; Major technological equipment (Primary circuit: Nuclear reactor [WWER 440 V230 type reactor];Steam generator; Reactor Coolant Pumps; Primary Circuit Auxiliary Systems. Secondary circuit: Turbine generators, Nuclear power plant electrical equipment; power plant control) and technical data

78 FR 77508 - Duke Energy Carolinas, LLC; William States Lee III Nuclear Station, Units 1 and 2; Combined...

Science.gov (United States)

2013-12-23

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-018 and 52-019; NRC-2008-0170] Duke Energy Carolinas, LLC; William States Lee III Nuclear Station, Units 1 and 2; Combined Licenses Application Review AGENCY: Nuclear Regulatory Commission. ACTION: Final environmental impact statement; availability...

The experience in the Cernavoda Unit 1 operation - a stimulating argument for future nuclear power development in Romania

International Nuclear Information System (INIS)

Rotaru, I.; Bucur, I.; Galeriu, A.C.; Budan, O.

1999-01-01

The Romanian nuclear program has been developed based on the option for CANDU type reactors. At the beginning, this program was unrealistically conceived and its management was inappropriate. The program was reconsidered in 1990 and the management policy and organization structure were also adapted accordingly. The paper presents, in the first part, the actual organization structure, adapted for the execution of the current and future activities, related to the nuclear power program. The performance achieved by Cernavoda Unit 1 constitutes the main part of the paper. The performances described demonstrate that the Cernavoda Unit 1 is a success and the Romania's electricity needs are satisfied in a proportion of about 12% by the nuclear power. The paper also presents a general view on Cernavoda Unit 2 perspectives. The essential conclusion of the paper is that the continuation of the nuclear program appears to be a logical option, generally accepted in Romania, limited only by financial restraints. (author)

A study on the optimal replacement periods of digital control computer's components of Wolsung nuclear power plant unit 1

International Nuclear Information System (INIS)

Mok, Jin Il; Seong, Poong Hyun

1993-01-01

Due to the failure of the instrument and control devices of nuclear power plants caused by aging, nuclear power plants occasionally trip. Even a trip of a single nuclear power plant (NPP) causes an extravagant economical loss and deteriorates public acceptance of nuclear power plants. Therefore, the replacement of the instrument and control devices with proper consideration of the aging effect is necessary in order to prevent the inadvertent trip. In this paper we investigated the optimal replacement periods of the control computer's components of Wolsung nuclear power plant Unit 1. We first derived mathematical models of optimal replacement periods to the digital control computer's components of Wolsung NPP Unit 1 and calculated the optimal replacement periods analytically. We compared the periods with the replacement periods currently used at Wolsung NPP Unit 1. The periods used at Wolsung is not based on mathematical analysis, but on empirical knowledge. As a consequence, the optimal replacement periods analytically obtained and those used in the field show a little difference. (Author)

Interface between Core/TH Model and Simulator for OPR1000

Energy Technology Data Exchange (ETDEWEB)

Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk; Lee, Seung Ho; Suh, Jung Kwan [KEPRI, Daejeon (Korea, Republic of)

2009-05-15

OPR1000 simulator for ShinKori-Unit 1, which will be operated at 2815MWt of thermal core power, is being developed while the ShinKori-Unit 1 and 2 is being built. OPR1000 simulator adopted the RELAP5 R/T code, which is the adaptation of RELAP5 and NESTLE codes to run in real-time mode with graphical visualization, to model Nuclear Steam Supply System (NSSS) Thermal-Hydraulics (TH) and Reactor Core. The RELAP5 is an advanced, best estimate, reactor TH simulation code developed at Idaho National Engineering and Environment Laboratory(INEEL) and the NESTLE is a true two-energy group neutronics code that computes the neutron flux and power for each node at every time step. As a simulator environment, the 3KEYMASTER{sup TM}, a commercial environment tool of WSC is used.

Interface between Core/TH Model and Simulator for OPR1000

International Nuclear Information System (INIS)

Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk; Lee, Seung Ho; Suh, Jung Kwan

2009-01-01

OPR1000 simulator for ShinKori-Unit 1, which will be operated at 2815MWt of thermal core power, is being developed while the ShinKori-Unit 1 and 2 is being built. OPR1000 simulator adopted the RELAP5 R/T code, which is the adaptation of RELAP5 and NESTLE codes to run in real-time mode with graphical visualization, to model Nuclear Steam Supply System (NSSS) Thermal-Hydraulics (TH) and Reactor Core. The RELAP5 is an advanced, best estimate, reactor TH simulation code developed at Idaho National Engineering and Environment Laboratory(INEEL) and the NESTLE is a true two-energy group neutronics code that computes the neutron flux and power for each node at every time step. As a simulator environment, the 3KEYMASTER TM , a commercial environment tool of WSC is used

78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Science.gov (United States)

2013-07-03

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 72-1004, 72-40, 50-269, 50-270, and 50-287; NRC-2013-0135] Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel Storage Installation; Environmental Assessment and Finding of No Significant Impact AGENCY: Nuclear...

78 FR 45575 - Duke Energy Carolinas, LLC; Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Science.gov (United States)

2013-07-29

... NUCLEAR REGULATORY COMMISSION [Docket Nos.: 72-1004, 72-40, 50-269, 50-270, 50-287; and NRC-2013- 0135] Duke Energy Carolinas, LLC; Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory Commission. ACTION: Exemption; issuance. SUMMARY: The NRC...

Integrated Plant Safety Assessment: Systematic Evaluation Program. Millstone Nuclear Power Station, Unit 1, Northeast Nuclear Energy Company, Docket No. 50-245. Final report

International Nuclear Information System (INIS)

1983-02-01

This report documents the review of the Millstone Nuclear Power Station, Unit 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit 1, is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license. This report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in November 1982

75 FR 14206 - FPL Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 1 and 2; Environmental Assessment...

Science.gov (United States)

2010-03-24

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-266 And 50-301; NRC-2010-0123 FPL Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 1 and 2; Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an Exemption, pursuant to...

Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

International Nuclear Information System (INIS)

Tanabe, Fumiya

2011-01-01

In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

A Nuclear Safety System based on Industrial Computer

International Nuclear Information System (INIS)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack

2011-01-01

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

A Nuclear Safety System based on Industrial Computer

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack [Korea Electric Power Corporation Engineering and Construction, Daejeon (Korea, Republic of)

2011-05-15

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

Internal event analysis of Laguna Verde Unit 1 Nuclear Power Plant. System Analysis

International Nuclear Information System (INIS)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

1993-01-01

The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis of Laguna Verde Unit 1 Nuclear Power Plant , CNSNS-TR-004, in five volumes. The reports are organized as follows: CNSNS-TR-004 Volume 1: Introduction and Methodology. CNSNS-TR-004 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR-004 Volume 3: System Analysis. CNSNS-TR-004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR-004 Volume 5: Appendices A, B and C. This volume presents the results of the system analysis for the Laguna Verde Unit 1 Nuclear Power Plant. The system analysis involved the development of logical models for all the systems included in the accident sequence event tree headings, and for all the support systems required to operate the front line systems. For the Internal Event analysis for Laguna Verde, 16 front line systems and 5 support systems were included. Detailed fault trees were developed for most of the important systems. Simplified fault trees focusing on major faults were constructed for those systems that can be adequately represent,ed using this kind of modeling. For those systems where fault tree models were not constructed, actual data were used to represent the dominant failures of the systems. The main failures included in the fault trees are hardware failures, test and maintenance unavailabilities, common cause failures, and human errors. The SETS and TEMAC codes were used to perform the qualitative and quantitative fault tree analyses. (Author)

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-02-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

Study on the Operating Strategy of HVAC Systems for Nuclear Decommissioning Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung-hwan; Han, Sung-heum; Lee, Jae-gon [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

According as Kori nuclear power plant unit 1 was determined to be defueled in 2017, various studies on nuclear plant decommissioning have been performed. In nuclear decommissioning plant, HVAC systems with large fan and electric coil have to be operated for long periods of time to support various types of work from defueled phase to final dismantling phase. So, in view of safety and utility costs, their overall operating strategy need to be established prior to defueled phase. This study presents HVAC system operating strategy at each decommissioning phase, that is, defueled plant operating phase, SSCs(systems, structures, components) decontamination and dismantling phases. In defueled plant operating phase, all fuel assemblies in reactor vessel are transferred to spent fuel pool(SFP) permanently. In defueled plant operation phase, reduction of the operating system trains is more practicable than the introduction of new HVAC components with reduced capacity. And, based on the result of the accident analyses for this phase, HVAC design bases such as MCR habitability requirement can be mitigated. According to these results, associated SSCs also can be downgraded. In similar approach, at each phase of plant decommissioning, proper inside design conditions and operating strategies should be re-established.

Present and future of Korean nuclear power

International Nuclear Information System (INIS)

Min, K-H

2014-01-01

'Full text:' The Korean nuclear power industry has devoted itself to technological development and self-reliance over the last 30 years since Kori unit 1, the first nuclear power plant commenced its commercial operation in 1978. As a result of such efforts and accumulated experiences, the Korean nuclear power industry has developed the OPR 1000 and APR 1400 units and is almost completing the development of the APR+ as a 1,500MW class reactor with its own technologies of design and manufacturing. Also, the Korean nuclear power industry has been able to build a strong supply chain from engineering, manufacturing, construction, and fuel supply, to operation and maintenance. At present, Korea is operating 23 commercial power reactors with a total installed capacity of 20,716 MW, accounting for 25 percent of the installed capacity and one third of the nation's total electricity generation. Also, the share of nuclear power generation capacity will be 29 percent by 2035 in the Long Term Energy Development Plan and 43 GW of nuclear energy capacity will be needed. Thanks to nuclear power generation as an essential driving force, Korea has been able to supply cheap and stable electricity. However, amid the growing public concerns about nuclear safety after the Fukushima accident, the Korean government and related organizations are exerting its utmost effort in all areas, for example, enhancing nuclear safety and safety culture, carrying out management innovation, and communicating with the public in order to enhance transparency. Also, the Korean government launched the Public Engagement Commission on spent nuclear fuel (SNF) management in 2013, which is tasked to initiate public consultation & discussion and submit recommendation to government after in-depth review and analysis on SNF management options by the end of 2014. Nuclear power has become very essential part of national economy in Korea because Korea has virtually no indigenous energy resources and

Present and future of Korean nuclear power

Energy Technology Data Exchange (ETDEWEB)

Min, K-H [Korea Atomic Industrial Forum, Inc., Seoul (Korea, Republic of)

2014-07-01

'Full text:' The Korean nuclear power industry has devoted itself to technological development and self-reliance over the last 30 years since Kori unit 1, the first nuclear power plant commenced its commercial operation in 1978. As a result of such efforts and accumulated experiences, the Korean nuclear power industry has developed the OPR 1000 and APR 1400 units and is almost completing the development of the APR+ as a 1,500MW class reactor with its own technologies of design and manufacturing. Also, the Korean nuclear power industry has been able to build a strong supply chain from engineering, manufacturing, construction, and fuel supply, to operation and maintenance. At present, Korea is operating 23 commercial power reactors with a total installed capacity of 20,716 MW, accounting for 25 percent of the installed capacity and one third of the nation's total electricity generation. Also, the share of nuclear power generation capacity will be 29 percent by 2035 in the Long Term Energy Development Plan and 43 GW of nuclear energy capacity will be needed. Thanks to nuclear power generation as an essential driving force, Korea has been able to supply cheap and stable electricity. However, amid the growing public concerns about nuclear safety after the Fukushima accident, the Korean government and related organizations are exerting its utmost effort in all areas, for example, enhancing nuclear safety and safety culture, carrying out management innovation, and communicating with the public in order to enhance transparency. Also, the Korean government launched the Public Engagement Commission on spent nuclear fuel (SNF) management in 2013, which is tasked to initiate public consultation & discussion and submit recommendation to government after in-depth review and analysis on SNF management options by the end of 2014. Nuclear power has become very essential part of national economy in Korea because Korea has virtually no indigenous energy resources and

76 FR 79228 - Combined Licenses at William States Lee III Nuclear Station Site, Units 1 and 2; Duke Energy...

Science.gov (United States)

2011-12-21

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-018 and 52-019; NRC-2008-0170] Combined Licenses at William States Lee III Nuclear Station Site, Units 1 and 2; Duke Energy Carolinas, LLC AGENCY: Nuclear.... SUMMARY: Notice is hereby given that the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Army Corps...

75 FR 13606 - Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

Science.gov (United States)

2010-03-22

... NUCLEAR REGULATORY COMMISSION [Docket Nos. STN 50-528, STN 50-529, and STN 50-530; NRC-2010-0114] Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3; Environmental...-74, issued to Arizona Public Service Company (APS, the licensee), for operation of the Palo Verde...

The Initial Test Programme Features for the Advanced Korean NPPs (Shin–Kori NPP Units 3&4)

International Nuclear Information System (INIS)

Sul, Y.S.

2015-01-01

Korea has developed the Advanced Power Reactor 1400 (APR1400), an evolutionary pressurised water reactor and has obtained the standard design approval in 2002. As of 2014, eight nuclear power plants (NPPs) are in preparation for operation or under construction, four in Korea (SKN 3&4, SUN 1&2) and four in UAE (BNPP 1, 2, 3&4), and four NPPs are in planning in Korea (SKN 5&6, SUN 3&4). Especially, SKN Units 3&4 are the first construction NPPs for the APR1400 and are currently in the final stage to get Operating Licence (OL). The initial test programme for NPPs begins as systems and components are turned over to the startup organization and ends with completion of the Power Ascension Tests (PATs). For SKN Unit 3, the Pre-core Hot Functional Testing has been successfully completed and the initial fuel loading would proceed after getting OL. The SKN Unit 4 is preparing for the CHT (Cold Hydrostatic Test). The SKN 3&4 has many new and advanced design features and so has developed the test programmes to demonstrate that those advanced design features can be safely operated and the performance levels can be maintained in accordance with approved safety requirements. Among those test programs, the examples of test programs, newly introduced ones compared to the previous NPPs are as follows; – SIT (Safety Injection Tank) Blowdown Test with FD (Fluidic Device) – IRWST (In-Containment Refueling Water Storage Tank) In-Plant Test – POSRV (Pilot Operated Safety Relief Valve) Test – Low Power Physics Testing and PAT considering FOAK (First Of A Kind) unit – Safe Shutdown Test with DCS (Distrubuted Control System) Fail and CMF (Common Mode Failure) for Safety Instrumentation and Control System. (author)

Upgrade of Control and Protection System of the Ignalina Nuclear Power Plant Units 1 and 2

International Nuclear Information System (INIS)

Wright, Ronald E.; Fletcher, Norman; Sidnev, Victor E.; Bickel, John H.; Vianello, Aldo; Pearsall, Raymond D.

2003-01-01

The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy

A Study on the Main Steam Safety Valve Opening Mechanism by Flashing on NPPs

International Nuclear Information System (INIS)

Kim, Bae Joo

2009-01-01

A safety injection event happened by opening of the Main Steam Safety Valve at Kori unit 1 on April 16, 2005. The safety valves were opened at the lower system pressure than the valve opening set point due to rapid system pressure drop by opening of the Power Operated Relief Valve installed at the upstream of the Main Steam System. But the opening mechanism of safety valve at the lower set point pressure was not explained exactly. So, it needs to be understood about the safety valve opening mechanism to prevent a recurrence of this kind of event at a similar system of Nuclear Power Plant. This study is aimed to suggest the hydrodynamic mechanism for the safety valve opening at the lower set point pressure and the possibility of the recurrence at similar system conditions through document reviewing for the related previous studies and Kori unit 1 event

Summary of plant life management evaluation for Onagawa Nuclear Power Station Unit-1

International Nuclear Information System (INIS)

Nodate, Kazumi

2014-01-01

The Onagawa Nuclear Power Station Unit-1 (Onagawa NPS-1) began commercial operation on June 1, 1984, and has reached 30-year from starting of operation on June of 2014. To that end, we implemented the Plant Life Management (PLM) evaluation for Onagawa NPS-1 as our first experience. We decided on a Long-term Maintenance Management Policy from result of the evaluation, and then applied the Safety-Regulations change approval application on November 6, 2013 and its correcting application on April 16, 2014. Our application was approved on May 21, 2014 through investigation by the Nuclear Regulatory Agency. Also at implementation of the PLM evaluation, we considered effects of the Great East Japan Earthquake that occurred on March 11, 2011 against ageing phenomena. In this paper, we introduce summary of PLM evaluation for Onagawa NPS-1 and the evaluation that considered effects of the Great East Japan Earthquake. (author)

75 FR 75704 - Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 And 2); Notice of...

Science.gov (United States)

2010-12-06

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-275-LR; 50-323-LR] Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 And 2); Notice of Appointment of Adjudicatory Employee... Seismologist, Office of Nuclear Material Safety and Safeguards, has been appointed as a Commission adjudicatory...

Probabilistic fire risk assessment for Koeberg Nuclear Power Station Unit 1

International Nuclear Information System (INIS)

Grobbelaar, J.F.; Foster, N.A.S.; Luesse, L.J.

1995-01-01

A probabilistic fire risk assessment was done for Koeberg Nuclear Power Station Unit 1. Areas where fires are likely to start were identified. Equipment important to safety, as well as their power and/or control cable routes were identified in each fire confinement sector. Fire confinement sectors where internal initiating events could be caused by fire were identified. Detection failure and suppression failure fault trees and event trees were constructed. The core damage frequency associated with each fire confinement sector was calculated, and important fire confinement sectors were identified. (author)

Final environmental statement for Shoreham Nuclear Power Station, Unit 1: (Docket No. 50-322)

International Nuclear Information System (INIS)

1977-10-01

The proposed action is the issuance of an Operating License to the Long Island Lighting Company (LILCO) for the startup and operation of the Shoreham Nuclear Power Station, Unit 1 (the plant) located on the north shore of Long Island, the State of New York, County of Suffolk, in the town of Brookhaven. The Shoreham station will employ a boiling-water reactor (BWR), which will operate at a thermal output of 2436 MW leading to a gross output of 846 MWe and a net output of about 820 MWe. The unit will be cooled by once-through flow of water from the Long Island Sound. One nuclear unit with a net capacity of 820 MWe will be added to the generating resources of the Long Island Lighting Company. This will have a favorable effect on reserve margins and provide a cost savings of approximately $62.1 million (1980 dollars) in production costs in 1980 if the unit comes on line as scheduled; additional cost savings will be realized in subsequent years. Approximately 100 acres (40 hectares) of the 500-acre (202-hectare) site of rural (mostly wooded) land owned by the applicant have been cleared. Most of this will be unavailable for other uses during at least the 40-year life of the plant. No offsite acreage has been or will be cleared. Land in the vicinity of the site has undergone some residential development that is typical for all of this area of Long Island. The operation of Shoreham Unit 1 will have insignificant impacts on this and other types of land uses in the vicinity of the site. 33 figs., 56 tabs

A pilot application of risk-informed methods to establish inservice inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station. Revision 1

International Nuclear Information System (INIS)

Vo, T.V.; Phan, H.K.; Gore, B.F.; Simonen, F.A.; Doctor, S.R.

1997-02-01

As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest National Laboratory has developed risk-informed approaches for inservice inspection plans of nuclear power plants. This method uses probabilistic risk assessment (PRA) results to identify and prioritize the most risk-important components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot application of this methodology. This report, which incorporates more recent plant-specific information and improved risk-informed methodology and tools, is Revision 1 of the earlier report (NUREG/CR-6181). The methodology discussed in the original report is no longer current and a preferred methodology is presented in this Revision. This report, NUREG/CR-6181, Rev. 1, therefore supersedes the earlier NUREG/CR-6181 published in August 1994. The specific systems addressed in this report are the auxiliary feedwater, the low-pressure injection, and the reactor coolant systems. The results provide a risk-informed ranking of components within these systems

Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

Energy Technology Data Exchange (ETDEWEB)

Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-07-01

During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

Safety-evaluation report related to operation of McGuire Nuclear Station, Units 1 and 2. Docket Nos. 50-369 and 50-370

International Nuclear Information System (INIS)

1983-05-01

This report supplements the Safety Evaluation Report Related to the Operation of McGuire Nuclear Station, Units 1 and 2 (SER (NUREG-0422)) issued in March 1978 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, as applicant and owner, for licenses to operate the McGuire Nuclear Station, Units 1 and 2 (Docket Nos. 50-369 and 50-370). The facility is located in Mecklenburg County, North Carolina, about 17 mi north-northwest of Charlotte, North Carolina. This supplement provides information related to issuance of a full-power authorization for Unit 2. The staff concludes that the McGuire Nuclear Station can be operated by the licensee without endangering the health and safety of the public

System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

International Nuclear Information System (INIS)

Ellis, R.J.

2000-01-01

The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

NSSS design and cycle 1 operating history data for Arkansas Nuclear One, Unit-2. Final report

International Nuclear Information System (INIS)

Gagne, P.A.

1981-03-01

This report contains design and cycle 1 operating data for the Arkansas Nuclear One, Unit-2 nuclear steam supply system. The design data include descriptions of the reactor core, reactor coolant system, and control systems which are a part of the nuclear steam supply system. Operating history data are provided for the period of December 1978 through January 1980. The most important operating history data provided include reactor power, cumulative fuel burnup, control rod position, primary coolant temperature, and a series of power distribution state points

Nuclear development in the United States

International Nuclear Information System (INIS)

Brewer, S.

1983-01-01

The history of the nuclear development in the United States has been one of international cooperation relations so far. The United States is to offer the technical information on atomic energy utilization to foreign countries in exchange for the guarantee that they never attempt to have or develop nuclear weapons. Actually, the United States has supplied the technologies on nuclear fuel cycle and other related fields to enable other countries to achieve economical and social progress. The Department of Energy clarified the public promise of the United States regarding the idea of international energy community. The ratio of nuclear power generation to total electric power supply in the United States exceeded 12%, and will exceed 20% by 1990. Since 1978, new nuclear power station has not been ordered, and some of the contracted power stations were canceled. The atomic energy industry in the United States prospered at the beginning of 1970s, but lost the spirit now, mainly due to the institutional problems rather than the technical ones. As the policy of the government to eliminate the obstacles, the improvement of the procedure for the permission and approval, the establishment of waste disposal capability, the verification of fast breeder reactor technology and the promotion of commercial fuel reprocessing were proposed. The re-establishment of the United States as the reliable supplier of atomic energy service is the final aim. (Kako, I.)

75 FR 15745 - Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3...

Science.gov (United States)

2010-03-30

...] Arizona Public Service Company, et al. Palo Verde Nuclear Generating Station, Units 1, 2, and 3; Exemption 1.0 Background The Arizona Public Service Company (APS, the licensee) is the holder of Facility... Generating Station (PVNGS), Units 1, 2, and 3, respectively. The licenses provide, among other things, that...

Nuclear safety inspection in treatment process for SG heat exchange tubes deficiency of unit 1, TNPS

International Nuclear Information System (INIS)

Zhang Chunming; Song Chenxiu; Zhao Pengyu; Hou Wei

2006-01-01

This paper describes treatment process for SG heat exchange tubes deficiency of Unit 1, TNPS, nuclear safety inspection of Northern Regional Office during treatment process for deficiency and further inspection after deficiency had been treated. (authors)

Development of a steam generator lancing system

International Nuclear Information System (INIS)

Jeong, Woo-Tae; Kim, Seok-Tae; Hong, Sung-Yull

2006-01-01

It is recommended to clean steam generators of nuclear power plants during plant outages. Under normal operations, sludge is created and constantly accumulates in the steam generators. The constituents of this sludge are different depending on each power plant characteristics. The sludge of the Kori Unit 1 steam generator, for example, was found to be composed of 93% ferrous oxide, 3% carbon and 1% of silica oxide and nickel oxide each. The research to develop a lancing system that would remove sludge deposits from the tubesheet of a steam generator was started in 1998 by the Korea Electric Power Research Institute (KEPRI) of the Korea Electric Power Corporation (KEPCO). The first commercial domestic lancing system in Korea, and KALANS-I Lancing System, was completed in 2000 for Kori Unit 1 for cleaning the tubesheet of its Westinghouse Delta-60 steam generator. Thereafter, the success of the development and site implementation of the KALANS-I lancing system for YGN Units 1 and 2 and Ulchin Units 3 and 4 was also realized in 2004 for sludge removal at those sites. The upper bundle cleaning system for Westinghouse model F steam generators is now under development

Comparison of the Radionuclides Dispersion at the UAE Barakah Site with that at the ROK Shin-Kori Site - Comparison of the radionuclides dispersion in Barakah site with that in Shin-Kori site

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung-yeop; Lee, Kun Jai; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of); Beeley, Philip A. [Khalifa University of Science, Technology and Research, P.O. Box 127788, Abu Dhabi (United Arab Emirates)

2014-07-01

In order to understand the characteristics of atmospheric dispersion of radionuclides in the desert environment of Barakah site in UAE, comparison research with the results of other environments could be an appropriate way to facilitate it. Shin-Kori site is the proper comparison target because same reactor type of APR1400 with that in Barakah site is under construction. Hypothetical accident scenario was considered and accident source term which had been developed in previous research has been applied as releasing source. After reviewing several computation codes, ADMS5 has been selected as an atmospheric dispersion modeling tool which is installing advanced Gaussian plum model and plentiful options. The climate data of both Barakah and Shin-Kori were acquired and the environments of both sites have been simulated considering wind speed, wind direction, temperature, humidity, ground surface roughness and etc. Near field final human doses on the maps have been schematised regarding statistical meteorological data of both sites and dose conversion factors from the publications of ICRP and federal guidance report of EPA. The results of this research are expected to enhance the understanding about differences between two environments which have same reactor type and to improve the comprehension of desert environment of Barakah site as well. Applying different dose conversion factors to Barakah site considering the desert biosphere could be further study to obtain more accurate results. (authors)

Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes

International Nuclear Information System (INIS)

Huerta B, A.; Lopez M, R.

1995-01-01

This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and.analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes

International Nuclear Information System (INIS)

Huerta B, A.; Lopez M, R.

1995-01-01

This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

Conformance to Regulatory Guide 1.97, Perry Nuclear Power Plant, Unit Nos. 1 and 2 (Docket Nos. 50-440 and 50-441)

International Nuclear Information System (INIS)

Udy, A.C.

1985-03-01

This EG and G Idaho, Inc., report provides a review of the submittals for Regulatory Guide 1.97, Revision 2, for the Perry Nuclear Power Plant, Unit Nos. 1 and 2. Any exception to the guidelines of Regulatory Guide 1.97 are evaluated

Architecture of nuclear power units

International Nuclear Information System (INIS)

Malaniuk, B.

1981-01-01

Nuclear units with circulation cooling using cooling towers are dominating points of the landscape. The individual cooling towers or pairs of cooling towers should be situated in the axes of double units and should also linearly be arranged, rhythmically in the respective zone. Examples are shown of the architectural designs of several nuclear power plants in the USA, the UK, the USSR, France, the FRG and Italy. (H.S.)

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

Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

International Nuclear Information System (INIS)

1983-12-01

Supplement 20 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Unit 1 and Unit 2 (Docket Nos. 50-275 and 50-323), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the verification effort for Diablo Canyon Unit 1 that was performed between November 1981 and the present in response to Commission Order CLI-81-30 and an NRC letter of November 19, 1981 to the licensee. Specifically, Supplement 20 addresses those issues and other matters identified in Supplements 18 and 19 that must be resolved prior to Unit 1 achieving criticality and operating at power levels up to 5% of rated full power. This SER Supplement applies only to Diablo Canyon Unit 1

76 FR 81994 - UniStar Nuclear Energy; Combined License Application for Calvert Cliffs Nuclear Power Plant, Unit...

Science.gov (United States)

2011-12-29

... NUCLEAR REGULATORY COMMISSION [Docket No. 52-016; NRC-2008-0250] UniStar Nuclear Energy; Combined License Application for Calvert Cliffs Nuclear Power Plant, Unit 3; Exemption 1.0 Background: UniStar Nuclear Energy (UNE) submitted to the U.S. Nuclear Regulatory Commission (NRC or the Commission ) a...

Detection of sensor degradation using K-means clustering and support vector regression in nuclear power plant

International Nuclear Information System (INIS)

Seo, Inyong; Ha, Bokam; Lee, Sungwoo; Shin, Changhoon; Lee, Jaeyong; Kim, Seongjun

2011-01-01

In a nuclear power plant (NPP), periodic sensor calibrations are required to assure sensors are operating correctly. However, only a few faulty sensors are found to be rectified. For the safe operation of an NPP and the reduction of unnecessary calibration, on-line calibration monitoring is needed. In this study, an on-line calibration monitoring called KPCSVR using k-means clustering and principal component based Auto-Associative support vector regression (PCSVR) is proposed for nuclear power plant. To reduce the training time of the model, k-means clustering method was used. Response surface methodology is employed to efficiently determine the optimal values of support vector regression hyperparameters. The proposed KPCSVR model was confirmed with actual plant data of Kori Nuclear Power Plant Unit 3 which were measured from the primary and secondary systems of the plant, and compared with the PCSVR model. By using data clustering, the average accuracy of PCSVR improved from 1.228×10 -4 to 0.472×10 -4 and the average sensitivity of PCSVR from 0.0930 to 0.0909, which results in good detection of sensor drift. Moreover, the training time is greatly reduced from 123.5 to 31.5 sec. (author)

Physical protection of nuclear operational units

International Nuclear Information System (INIS)

1981-07-01

The general principles of and basic requirements for the physical protection of operational units in the nuclear field are established. They concern the operational units whose activities are related with production, utilization, processing, reprocessing, handling, transport or storage of materials of interest for the Brazilian Nuclear Program. (I.C.R.) [pt

A Study on the Construct Validity of Safety Culture Oversight Model for Nuclear Power Operating Organization

International Nuclear Information System (INIS)

Jung, Su Jin; Choi, Young Sung; Oh, Jang Jin

2015-01-01

In Korea, the safety policy statement declared in 1994 by government stressed the importance of safety culture and licensees were encouraged to manage and conduct their self-assessments. A change in regulatory position about safety culture oversight was made after the event of SBO cover-up in Kori unit 1 and several subsequent falsification events. Since then KINS has been developing licensee's safety culture oversight system including conceptual framework of oversight, prime focus area for oversight, and specific details on regulatory expectations, all of which are based on defence-in-depth (DiD) safety enhancement approach. Development and gathering of performance data which is related to actual 'safety' of nuclear power plant are needed to identify the relationship between safety culture and safety performance. Authors consider this study as pilot which has a contribution on verifying the construct validity of the model and the effectiveness of survey based research. This is the first attempt that the validity of safety culture oversight model has been investigated with empirical data obtained from Korean nuclear power operating organization

A Study on the Construct Validity of Safety Culture Oversight Model for Nuclear Power Operating Organization

Energy Technology Data Exchange (ETDEWEB)

Jung, Su Jin; Choi, Young Sung; Oh, Jang Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

In Korea, the safety policy statement declared in 1994 by government stressed the importance of safety culture and licensees were encouraged to manage and conduct their self-assessments. A change in regulatory position about safety culture oversight was made after the event of SBO cover-up in Kori unit 1 and several subsequent falsification events. Since then KINS has been developing licensee's safety culture oversight system including conceptual framework of oversight, prime focus area for oversight, and specific details on regulatory expectations, all of which are based on defence-in-depth (DiD) safety enhancement approach. Development and gathering of performance data which is related to actual 'safety' of nuclear power plant are needed to identify the relationship between safety culture and safety performance. Authors consider this study as pilot which has a contribution on verifying the construct validity of the model and the effectiveness of survey based research. This is the first attempt that the validity of safety culture oversight model has been investigated with empirical data obtained from Korean nuclear power operating organization.

Computation code TEP 1 for automated evaluation of technical and economic parameters of operation of WWER-440 nuclear power plant units

International Nuclear Information System (INIS)

Zadrazil, J.; Cvan, M.; Strimelsky, V.

1987-01-01

The TEP 1 program is used for automated evaluation of the technical and economic parameters of nuclear power plant units with WWER-440 reactors. This is an application program developed by the Research Institute for Nuclear Power Plants in Jaslovske Bohunice for the KOMPLEX-URAN 2M information system, delivered by the USSR to the V-2 nuclear power plants in Jaslovske Bohunice and in Dukovany. The TEP 1 program is written in FORTRAN IV and its operation has two parts. First the evaluation of technical and economic parameters of operation for a calculation interval of 10 mins and second, the control of the calculation procedure, follow-up on input data, determination of technical and economic parameters for a lengthy time interval, and data printout and storage. The TEP 1 program was tested at the first unit of the V-2 power plant and no serious faults appeared in the process of the evaluation of technical and economic parameters. A modification of the TEP 1 programme for the Dukovany nuclear power plant is now being tested on the first unit of the plant. (Z.M.)

Safety-evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322)

International Nuclear Information System (INIS)

1983-09-01

Supplement 4 (SSER 4) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1. Docket No. 50-322

International Nuclear Information System (INIS)

1983-02-01

Supplement No. 3 to the Safety Evaluation Report of Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have come to light since the previous supplement was issued

Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322)

International Nuclear Information System (INIS)

1989-04-01

Supplement 10 (SSER 10) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Technical specifications for Grand Gulf Nuclear Station, Unit 1 (Docket No. 50-416). Appendix A to License No. NPF-13

International Nuclear Information System (INIS)

1984-08-01

The Grand Gulf Nuclear Station, Unit 1 Technical Specifications were prepared by the US Nuclear Regulatory Commission to set forth the limits, operating conditions and other requirements applicable to a nuclear facility as set forth in Section 50.36 of 10 CFR part 50 for the protection of the health and safety of the public

Kozloduy nuclear power plant. Units 1-4. Status of safety improvements. Rev. 2

International Nuclear Information System (INIS)

1999-01-01

operation of the units. The plant current safety level analysis has been performed using IAEA analytical methodology according to 50-SG-O12 standard 'Periodic safety review of operational nuclear power plants'. The approach and criteria for acceptable safety level definition, developed by IAEA and presented in INSAG-8 ' was used for analysis performance. Based on this analysis a set of activities was developed, to ensures further plant operation with the necessary safety level. The measures were combined in a program called Complex program for modernization of units 1-4. The implementation of the program is foreseen for a period of four next fuel cycles and started at the beginning of 1998. In respond to the requirements for the content of this paper a detailed description of the current status of resolving of the safety issues, classified by IAEA in TECDOC 640 is presented. The whole process of safety evaluation, short and long term safety improvements presents in a systematical manner the efforts of the Government of Bulgaria, NEK Ltd and KNPP to operate these units with due respect of their nuclear safety responsibility according to Nuclear Safety Convention signed at Vienna in 1994 and ratified by Bulgarian Parliament in 1995

Nuclear power plant V-1

International Nuclear Information System (INIS)

1999-01-01

In this leaflet the short history of commissioning of Bohunice V-1 NPP is reviewed (beginning of construction 24 April 1972; First controlled reactor power, Reactor Unit 1 (RU1): 27 November 1978, Reactor Unit 2 (RU2): 15 March 1980; Connection to the grid: RU1 17 December 1978, RU2 26 March 1980; Commercial operation: RU1 1 April 1980, RU2 7 January 1981. The scheme of the nuclear reactor WWER 440/V230 is depicted. The major technological equipment (primary circuit, nuclear reactor, steam generators, reactor coolant pumps, primary circuit auxiliary systems, secondary circuit, turbine generators, NPP electrical equipment, and power plant control) are described. Technical data of the Bohunice V-1 NPP are presented

The software testing of PPS for shin Ulchin nuclear power plant units 1 and 2

International Nuclear Information System (INIS)

Kang, Dong Pa; Park, Cheol Lak; Cho, Chang Hui; Sohn, Se Do; Baek, Seung Min

2012-01-01

The testing of software (S/W) is the process of analyzing a software item to detect the differences between existing and required conditions to evaluate the features of the software items. This paper introduces the S/W testing of Plant Protection System (PPS), as a safety system which actuate Reactor Trip (RT) and Engineered Safety Features (ESF) for Shin Ulchin Nuclear Power Plant Units 1 and 2 (SUN 1 and 2)

Safety evaluation report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 1

International Nuclear Information System (INIS)

1984-06-01

This report, Supplement No. 1 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicant) for license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report

Analysis of safety information for nuclear power plants and development of source term estimation program

International Nuclear Information System (INIS)

Kim, Tae Woon; Choi, Seong Soo; Park, Jin Hee

1999-12-01

Current CARE(Computerized Advisory System for Radiological Emergency) in KINS(Korea Institute of Nuclear Safety) has no STES(Source Term Estimation System) which links between SIDS(Safety Information Display System) and FADAS(Following Accident Dose Assessment System). So in this study, STES is under development. STES system is the system that estimates the source term based on the safety information provided by SIDS. Estimated source term is given to FADAS as an input for estimation of environmental effect of radiation. Through this first year project STES for the Kori 3,4 and Younggwang 1,2 has been developed. Since there is no CARE for Wolsong(PHWR) plants yet, CARE for Wolsong is under construction. The safety parameters are selected and the safety information display screens and the alarm logic for plant status change are developed for Wolsong Unit 2 based on the design documents for CANDU plants

Critical evaluation of the nonradiological environmental technical specifications. Volume 4. San Onofre Nuclear Generating Station, Unit 1

Energy Technology Data Exchange (ETDEWEB)

Adams, S.M.; Cunningham, P.A.; Gray, D.D.; Kumar, K.D.

1976-08-10

A comprehensive study of the data collected as part of the environmental Technical Specifications program for Unit 1 of the San Onofre Nuclear Generating Station (SONGS 1) was conducted for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The program included an analysis of the hydrothermal and ecological monitoring data collected during 1975. The hydrothermal analysis includes a discussion of models used in plume predictions prior to plant operation and an evaluation of the present hydrothermal monitoring program. The ecological evaluation was directed toward reviewing the strengths and weaknesses of the various sampling programs designed to monitor the planktonic, benthic, and nektonic communities inhabiting the inshore coastal area in the vicinity of San Onofre.

Critical evaluation of the nonradiological environmental technical specifications. Volume 4. San Onofre Nuclear Generating Station, Unit 1

International Nuclear Information System (INIS)

Adams, S.M.; Cunningham, P.A.; Gray, D.D.; Kumar, K.D.

1976-01-01

A comprehensive study of the data collected as part of the environmental Technical Specifications program for Unit 1 of the San Onofre Nuclear Generating Station (SONGS 1) was conducted for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The program included an analysis of the hydrothermal and ecological monitoring data collected during 1975. The hydrothermal analysis includes a discussion of models used in plume predictions prior to plant operation and an evaluation of the present hydrothermal monitoring program. The ecological evaluation was directed toward reviewing the strengths and weaknesses of the various sampling programs designed to monitor the planktonic, benthic, and nektonic communities inhabiting the inshore coastal area in the vicinity of San Onofre

Second unit scheduling concerns on a dual-unit nuclear project

International Nuclear Information System (INIS)

Block, H.R.; Mazzini, R.A.

1978-01-01

This paper explores the planning and scheduling problems of Unit 2 of the Susquehanna steam electric station. The causes of these problems and methods to avoid or mitigate their consequences are discussed. The Susquehanna steam electric station has two boiling water reactors rated at 1,100 MW each. Topics considered include cost factors, structures, equipment, engineering and home office, construction services, completion data phasing, work sequencing, structural dependences, and segregation. Substatial cost and schedule benefits can result if two nuclear units are designed and constructed as one integral station, and if maximum sharing of facilities and services between the units occurs. It is concluded that the cost benefits of highly integrated dual unit construction outweigh the schedule and logistical problems caused by that approach

Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

International Nuclear Information System (INIS)

1984-07-01

Supplement 27 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for a license to operate Diablo Canyon Nuclear Power Plant, Unit 1 (Docket No. 50-275), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses the revisions to the license conditions and to the Technical Specifications as they relate to Amendment 10 to Diablo Canyon, Unit 1 Facility Operating License, DPR-76

Technical evaluation of RETS-required reports for Rancho Seco Nuclear Generating Station, Unit 1 for 1983

International Nuclear Information System (INIS)

Magleby, E.H.; Young, T.E.

1985-01-01

A review of the reports required by Federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted during 1983 was performed. The periodic reports reviewed for the Rancho Seco Nuclear Generating Station, Unit 1 were the Semiannual Effluent Release Report, January 1, 1983 to June 30, 1983 and the Radiation Exposure, Environmental Protection, Effluent and Waste Disposal Report. The principal review guidelines were the plant's specific RETS which were based on NRC guidance given in NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants.'' The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414)

International Nuclear Information System (INIS)

1984-07-01

The report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for fuel loading and precriticality testing for Unit 1

Development of software for the microsimulator for the KO-RI nuclear power plant unit 2

International Nuclear Information System (INIS)

Seok, H.; No, H.C.; Cho, S.J.; Park, S.D.; Jun, H.Y.; Lee, Y.K.

1994-01-01

A workstation-based real-time simulator for two-loop pressurized water reactor plants is developed for classroom training in support of a full-scale simulator, on-site transient analysis, and engineering studies. The present simulator consists of three functional modules: plant module, graphic module, and man-machine interaction module. The plant module includes models for the core kinetics, reactor coolant system, steam generator, main steam line, balance of plant, and control and protection system. Each of the models is optimized to obtain the capability of real-time simulation. The graphic module is designed to provide the user with more information at a glance by dynamically displaying schematic diagrams of the systems, symbols indicating the operating status of each component, trend curves, and the main control room. As tools for the man-machine interface, the man-machine interaction model uses a color cathode ray tube monitor, a standard keyboard, and the mouse. The interactive communication module receives parameters from the user via the keyboard and mouse, and transfers them to the plant module so as to enable the user to perform his specific actions. This module provides the user with various initiating events (malfunctions and manual controls) through SYSTEM, CONTROL ROOM, and ACCIDENTS menus, and thus a wide range of nuclear steam supply system transients can be easily simulated. The FISA-2/WS is verified through comparisons with analytical solutions, separated tests and integral tests, and predictions by RETRAN-2 and RELAP5/MOD3

Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

International Nuclear Information System (INIS)

Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

1985-02-01

Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

Uslovi korišćenja plovnih dizalica za dizanje potonulih objekata na unutrašnjim plovnim putevima

OpenAIRE

Radojević M. Slobodan

2012-01-01

U radu se prikazuju uslovi korišćenja plovnih dizalica za dizanje potonulih plovnih i drugih objekata na unutrašnjim plovnim putevima. Prikazani su osnovni načini podizanja plovnim dizalicama i osnovni tehnički podaci sa proračunskim pojedinostima za predložen postupak dizanja. Ukazano je na značaj dizanja potonulih objekata i njihovog uklanjanja iz unutrašnjih plovnih puteva u Republici Srbiji.

Safety-Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2. Docket Nos. 50-416 and 50-417

International Nuclear Information System (INIS)

1983-05-01

Supplement 4 to the Safety Evaluation Report for Mississippi Power and Light Company, et. al., joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status on the resolution of those issues that required further evaluation before authorizing operation of Unit 1 above 5% rated power and other issues that were to be evaluated during the first cycle of power operation

Common cause failure analysis of the rodded scram system of the Arkansas Nuclear One-Unit 1 Plant

International Nuclear Information System (INIS)

Montague, D.F.; Campbell, D.J.; Flanagan, G.F.

1986-10-01

This study demonstrates the use of a formal method for common cause failure analysis in a reliability analysis of the Arkansas Nuclear One - Unit 1 rodded scram system. The scram system failure of interest is loss of capability of the system to shut the reactor down when required. The results of this analysis support the ATWS program sponsored by the US Nuclear Regulatory Commission. The methods used in this analysis support the NRC's Risk Methods Integration and Evaluation Program (RMIEP)

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414)

International Nuclear Information System (INIS)

1984-12-01

This report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc., as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for initial criticality and power ascension to full-power opertion for Unit 1

Nuclear Data Unit correspondents for the exchange of nuclear data information

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-04-15

This list serves as a basis for the distribution of documents originated by or for the International Nuclear Data Committee, and includes the names of all INDC members, liaison officers and correspondents of the Nuclear Data Unit. The IAEA Nuclear Data Unit tries to maintain this list up-to-date in order to facilitate an efficient interchange of information on nuclear data topics. The recipients of this list are encouraged to inform the Nuclear Data Unit of any corrections, additions and deletions deemed necessary. Because of the variety of document origins and functions, and the number of countries served by this list, each addressee is assigned a seven-letter distribution code, shown to the left of each name, which indicates the categories of documents he receives. The names appear under three categories, member states, international organizations and non-member states, and are listed in alphabetical order within each state or organization. The main list is followed by seven shorter lists, indicating the names of individuals in each distribution category, and the total number of individuals in each category.

Nuclear Data Unit correspondents for the exchange of nuclear data information

International Nuclear Information System (INIS)

1968-04-01

This list serves as a basis for the distribution of documents originated by or for the International Nuclear Data Committee, and includes the names of all INDC members, liaison officers and correspondents of the Nuclear Data Unit. The IAEA Nuclear Data Unit tries to maintain this list up-to-date in order to facilitate an efficient interchange of information on nuclear data topics. The recipients of this list are encouraged to inform the Nuclear Data Unit of any corrections, additions and deletions deemed necessary. Because of the variety of document origins and functions, and the number of countries served by this list, each addressee is assigned a seven-letter distribution code, shown to the left of each name, which indicates the categories of documents he receives. The names appear under three categories, member states, international organizations and non-member states, and are listed in alphabetical order within each state or organization. The main list is followed by seven shorter lists, indicating the names of individuals in each distribution category, and the total number of individuals in each category

United Campuses to Prevent Nuclear War: Nuclear War Course Summaries.

Science.gov (United States)

Journal of College Science Teaching, 1983

1983-01-01

Briefly describes 46 courses on nuclear war available from United Campuses to Prevent Nuclear War (UCAM). These courses are currently being or have been taught at colleges/universities, addressing effects of nuclear war, arms race history, new weapons, and past arms control efforts. Syllabi (with assignments/reading lists) are available from UCAM.…

Republic of Korea [National and regional programmes on the production of hydrogen using nuclear energy

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-03-15

The total primary energy consumption of the Republic of Korea in 2006 was 233 Mtoe (ranking ninth in the world), with 43% petroleum, 24% coal, 16% nuclear, 14% LNG, 2% renewables and 1% hydro. Energy consumption is expected to grow significantly in the future. The country lacks domestic energy resources and currently has to import 97% of its primary energy demand. The Republic of Korea is the sixth largest and fastest growing CO{sub 2} emitter of the OECD countries. The total installed electrical generation capacity is 61.4 GW(e), of which 17.5 GW(e) is from nuclear. As of 2006, 36% of the electricity was generated by nuclear, 38% by coal, 20% by LNG, 5% by petroleum and 1% by hydropower. The Republic of Korea is a small country with a high population density where the use of low-density renewable energies is limited and not a practicable solution. Commercial scale nuclear power generation started at the Kori-1 plant in 1978, and another 19 reactor units have since been built using a mixture of CANDU (4 reactors) and PWR (16 reactors) technologies. The total nuclear capacity amounts to 17.7 GW. Eight more plants are planned to come on-line in the period from 2010 to 2016, adding another 9.4 GW. According to the 'National Energy Basic Plan' of 2008, the share of nuclear in the primary energy should grow to 33% provided by 32 units. Nuclear power research in the Republic of Korea is very active with investigation into a variety of advanced reactors, including the Korea Atomic Energy Research Institute (KAERI) small system-integrated modular advanced reactor (SMART), a 330 MW(th) pressurized water reactor with integral steam generators and advanced safety features, and designed for generating electricity (up to 100 MW(e)) and/or for thermal applications such as seawater desalination. Other advanced reactor concepts under development are a liquid metal fast/transmutation reactor and a high temperature hydrogen generation design.

Integrated safety assessment report: Integrated Safety Assessment Program: Millstone Nuclear Power Station, Unit 1 (Docket No. 50-245): Draft report

International Nuclear Information System (INIS)

1987-04-01

The Integrated Safety Assessment Program (ISAP) was initiated in November 1984, by the US Nuclear Regulatory Commission to conduct integrated assessments for operating nuclear power reactors. The integrated assessment is conducted in a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. In addition, procedures will be established to allow for a periodic updating of the schedules to account for licensing issues that arise in the future. This report documents the review of Millstone Nuclear Power Station, Unit No. 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit No. 1, is one of two plants being reviewed under the pilot program for ISAP. This report indicates how 85 topics selected for review were addressed. This report presents the staff's recommendations regarding the corrective actions to resolve the 85 topics and other actions to enhance plant safety. The report is being issued in draft form to obtain comments from the licensee, nuclear safety experts, and the Advisory Committee for Reactor Safeguards (ACRS). Once those comments have been resolved, the staff will present its positions, along with a long-term implementation schedule from the licensee, in the final version of this report

Development of a Field Management Standard for Improving Human Factors

International Nuclear Information System (INIS)

Yun, Young Su; Son, Il Moon; Son, Byung Chang; Kwak, Hyo Yean

2009-07-01

This project is to develop a management guideline for improving human performances as a part of the Human Factors Management System of Kori unit 1 which is managing all of human factors items such as man-machine system interfaces, work procedures, work environments, and human reliabilities in nuclear power plants. Human factors engineering includes an human factors suitability analysis and improvement of human works, an analysis of accidents by human error, an improvement of work environment, an establishment of human factors management rules and a development of human resources to manage and perform those things consistently. For assisting these human factors engineering tasks, we developed human factors management guidelines, checklists and work procedures to be used in staffing, qualification, training, and human information requirements and workload. We also provided a software tool for managing the above items. Additionally, contents and an item pool for a human factors qualifying examination and training programs were developed. A procedures improvement and a human factors V and V on the Kori unit 1 have been completed as a part of this project, too

Experience and development of on-line BWR surveillance system at Onagawa nuclear power station unit-1

International Nuclear Information System (INIS)

Kishi, A.; Chiba, K.; Kato, K.; Ebata, S.; Ando, Y.; Sakamoto, H.

1986-01-01

ONAGAWA nuclear power station Unit-1 (Tohoku Electric Power Co.) is a BWR-4 nuclear power station of 524 MW electric power which started commercial operation in June 1984. To attain high reliability and applicability for ONAGAWA-1, Tohoku Electric Power Co. and Toshiba started a Research and Development project on plant surveillance and diagnosis from April 1982. Main purposes of this project are to: (1) Develop an on-line surveillance system and acquire its operating experience at a commercial BWR, (2) Assist in plant operation and maintenance by data acquisition and analysis, (3) Develop a new technique for plant surveillance and diagnosis. An outline of the project, operating experience gained from the on-line surveillance system and an introduction to new diagnosis techniques are reported in this paper. (author)

Conformance to Regulatory Guide 1.97, Joseph M. Farley Nuclear Plant, Unit Nos. 1 and 2 (Docket No. 50-348 and 50-364)

International Nuclear Information System (INIS)

Stoffel, J.W.

1985-05-01

This EG and G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit Nos. 1 and 2 of the Joseph M. Farley Nuclear Plant and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified

Status of nuclear engineering education in the United States

International Nuclear Information System (INIS)

Brown, G.J.

2000-01-01

Nuclear engineering education in the United States is reflective of the perceived health of the nuclear electric power industry within the country. Just as new commercial reactor orders have vanished and some power plants have shut down, so too have university enrollments shrunk and research reactors closed. This decline in nuclear trained specialists and the disappearance of the nuclear infrastructure is a trend that must be arrested and reversed if the United States is to have a workforce capable of caring for a nuclear power industry to not only meet future electric demand but to ensure that the over 100 existing plants, their supporting facilities and their legacy in the form of high level waste and facility clean-up are addressed. Additionally, the United States has an obligation to support and maintain its nuclear navy and other defence needs. And, lastly, if the United States is to have a meaningful role in the international use of nuclear power with regard to safety, non-proliferation and the environment, then it is imperative that the country continues to produce world-class nuclear engineers and scientists by supporting nuclear engineering education at its universities. The continued support of the federal government. and industry for university nuclear engineering and nuclear energy research and development is essential to sustain the nuclear infrastructure in the United States. Even with this support, and the continued excellent operation of the existing fleet of nuclear electric power plants, it is conceivable that nuclear engineering as an academic discipline may fall victim to poor communications and a tarnished public image. What is needed is a combination of federal and industrial support along with the creativity of the universities to expand their offerings to include more than power production. The objective is a positive message on careers in nuclear related fields, and recognition of the important role of nuclear energy in meeting the country

Technical specifications, Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322): Appendix ''A'' to License No. NPF-82

International Nuclear Information System (INIS)

1989-04-01

The Shoreham, Unit 1, Technical Specifications were prepared by the US Nuclear Regulatory Commission to set forth the limits, operating conditions, and other requirements applicable to a nuclear facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. 20 figs., 75 tabs

Outline of construction and facility features of Onagawa nuclear power station Unit No. 2

International Nuclear Information System (INIS)

Umimura, Yoshiharu; Tsunoda, Ryohei; Watanabe, Kazunori

1996-01-01

Tohoku Electric Power Company promotes development of various power sources to provide a stable supply of electricity in the future, and nuclear power takes a leading part. In August 1989, construction of Onagawa nuclear power plant Unit No. 2 (825MW) was started, following Unit No. 1 (524MW) which went on line in 1984 as Tohoku Electric's first nuclear power plant unit. Unit No. 2 began commercial operation in July 1995 through satisfactory construction work such as RPV hydraulic test in March 1994, fuel loading in October 1994, and various startup tests in each power stage. The design and construction of Unit No. 2 reflect construction and operation experience gained from Unit No. 1, and the latest technology, including that of the LWR Improvement and Standardization Program, was adopted to enhance facility reliability, improve operation and maintenance performance, and reduce worker dosage. Features of the facility, construction techniques, and a description of preoperation of Onagawa nuclear power plant Unit No. 2 are described in this paper. (author)

Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

International Nuclear Information System (INIS)

1991-06-01

Supplement 34 to the Safety Evaluation Report for the application by Pacific Gas and Electric Company (PG ampersand E) for licenses to operate Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2 (Docket Nos. 50-275 and 50-323, respectively) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement documents the NRC staff review of the Long-Term Seismic Program conducted by PG ampersand E in response to License Condition 2.C.(7) of Facility Operating License DPR-80, the Diablo Canyon Unit 1 operating license. 111 refs., 20 figs., 31 tabs

Nuclear Legislation in OECD and NEA Countries. Regulatory and Institutional Framework for Nuclear Activities - United Kingdom

International Nuclear Information System (INIS)

2003-01-01

This country profile provide comprehensive information on the regulatory and Institutional Framework governing nuclear activities as well as a detailed review of a full range of nuclear law topics, including: mining regime; radioactive substances; nuclear installations; trade in nuclear materials and equipment; radiation protection; radioactive waste management; non-proliferation and physical protection; transport; and nuclear third party liability. The profile is complemented by reproductions of the primary legislation regulating nuclear activities in the country. Content: I. General Regulatory Regime: 1. Introduction; 2. Mining Regime; 3. Radioactive Substances; 4. Nuclear Installations (Licensing and inspection, including nuclear safety; Protection of the environment against radiation effects; Emergency response); 5. Trade in Nuclear Materials and Equipment; 6. Radiation Protection; 7. Radioactive Waste Management; 8. Non-Proliferation and Physical Protection; 9. Transport; 10. Nuclear Third Party Liability; II. Institutional Framework: 1. Regulatory and Supervisory Authorities (Department of Trade and Industry - DTI; Secretary of State for Environment, Food and Rural Affairs and the Secretary of State for Health; Secretary of State for Transport; Secretary of State for Education); 2. Advisory Bodies (Medical Research Council - MRC; Nuclear Safety Advisory Committee; Radioactive Waste Management Advisory Committee); 3. Public and Semi-Public Agencies (United Kingdom Atomic Energy Authority - UKAEA; Health and Safety Commission and Executive - HSC/HSE; National Radiological Protection Board - NRPB; Environment Agencies; British Nuclear Fuels plc. - BNFL; Amersham International plc.; The National Nuclear Corporation Ltd. - NNC; United Kingdom Nirex Ltd.; Magnox Electric plc.; British Energy Generation Ltd.; Scottish Electricity Generator Companies; British Energy Generation Ltd.; Regional Electricity Companies in England and Wales)

Cernavoda NPP Unit 1 - a plant of several generations

International Nuclear Information System (INIS)

Rotaru, I.; Metes, M.; Anghelescu, M.S.

2000-01-01

Cernavoda NPP Unit 1, the first nuclear power unit in Romania, has a long and tormented history. It represents a rather unique case in Eastern Europe. The project started well before 1989 (the construction phase lasted 17 years and generations were involved in its completion), but it is effectively based on western technology (Candu). Meanwhile, the national nuclear program underwent many changes, affecting the lives and careers of Romanian nuclear professionals. Finally, on December 2 nd 1996, the unit began its c ommercial operation , being operated at its nominal power rating of 706 MW e . It now provides a reliable source of electricity for Romanian economy, supplying to the national grid about 10% of the country's average annual demand. The paper reflects some aspects related to the shift of generations during the project's development, including the present stage. The operational performances achieved 'in service' by Cernavoda NPP Unit 1 up to the end of 1999 , are also presented. Reference to the electrical energy production, performance indicators, production costs, nuclear safety, radiation protection, radioactive wastes, nuclear fuel consumption and nuclear fuel performances are included. Comparisons are performed with similar indicators reported by other worldwide nuclear power plants, in order to assess our results. (authors)

Science, society, and America's nuclear waste: Unit 3, The Nuclear Waste Policy Act

International Nuclear Information System (INIS)

1992-01-01

This is the 3rd unit, (The Nuclear Waste Policy Act) a four-unit secondary curriculum. It is intended to provide information about scientific and societal issues related to the management of spent nuclear fuel from generation of electricity at nuclear powerplants and high-level radioactive waste from US national defense activities. The curriculum, supporting classroom activities, and teaching materials present a brief discussion of energy and electricity generation, including that produced at nuclear powerplants; information on sources, amounts, location, and characteristics of spent nuclear fuel and high-level radioactive waste; sources, types and effects of radiation; US policy for managing and disposing of spent nuclear fuel and high-level radioactive waste and what other countries are doing; and the components of the nuclear waste management system

Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417). Supplement 6

International Nuclear Information System (INIS)

1984-08-01

Supplement 6 to the Safety Evaluation Report for Mississippi Power and Light Company et al. joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the NRC staff's evaluation of open items from previous supplements and Technical Specification changes required before authorizing operation of Unit 1 above 5% of rated power

75 FR 44292 - Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2; Notice of...

Science.gov (United States)

2010-07-28

... and DPR-60] Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2... assessment, and behavioral observation) of the unescorted access authorization program when making the... under consideration to determine whether it met the criteria established in NRC Management Directive (MD...

Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 30

International Nuclear Information System (INIS)

1985-04-01

Supplement 30 to the Safety Evaluation Report for the application by the Pacific Gas and Electric Company (PG and E) to operate the Diablo Canyon Nuclear Power Plant - Unit 2 (Docket No. 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. SSER 30 reports on the staff's technical review and evaluation of the design and analysis of Unit 2 piping systems and pipe supports. The staff effort includes an evaluation of PG and E's treatment of issues raised during the Unit 1 design verification, actions resulting from low power License Condition 2.C.(11) in the Unit 1 low power license DPR-76 and the Unit 2 applicability and resolution of certain allegations related to piping and supports

Calculation of DPA in the Reactor Internal Structural Materials of Nuclear Power Plant

International Nuclear Information System (INIS)

Kim, Yong Deong; Lee, Hwan Soo

2014-01-01

The embrittlement is mainly caused by atomic displacement damage due to irradiations with neutrons, especially fast neutrons. The integrity of the reactor internal structural materials has to be ensured over the reactor life time, threatened by the irradiation induced displacement damage. Accurate modeling and prediction of the displacement damage is a first step to evaluate the integrity of the reactor internal structural materials. Traditional approaches for analyzing the displacement damage of the materials have relied on tradition model, developed initially for simple metals, Kinchin and Pease (K-P), and the standard formulation of it by Norgett et al. , often referred to as the 'NRT' model. An alternative and complementary strategy for calculating the displacement damage is to use MCNP code. MCNP uses detailed physics and continuous-energy cross-section data in its simulations. In this paper, we have performed the evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling and compared with predictions results of displacement damage using the classical NRT model. The evaluation of the displacement damage of the reactor internal structural materials in Kori NPP unit 1 using detailed Monte Carlo modeling has been performed. The maximum value of the DPA rate was occurred at the baffle side of the reactor internal where the node has the maximum neutron flux

The Bohunice NPP V-1 units nuclear safety upgrading

International Nuclear Information System (INIS)

Mlcuch, M.

2000-01-01

Safety upgrading and operational reliability improvement was carried out by the Bohunice NPP V-1 staff continuously since the plant commissioning. By now, more than 1200 minor or major modifications have been implemented, either by the NPP maintenance staff or by the contractors. Based on findings of safety assessment missions invited by Bohunice NPP in 1990 - 1991, the Czecho-slovak Nuclear Regulatory Authority (CSKAE) issued the decision No. 5/91 of 81 safety upgrading measures to be taken in different areas. These improvements are referred to as the 'Small Reconstruction of the Bohunice V-1 NPP'. Realization of measures during Small reconstruction of the Bohunice NPP V-1 became a power plant, which further operation is acceptable from safety point of view, but it is also necessary further safety improvement. During the period of the Small Reconstruction the development of a Safety Report for the Gradual reconstruction has been completed. Based on this report the SR Nuclear Regulatory Authority issued the Decision No. 1/94, in which requires 59 upgrading measures in different areas to be addressed. The development of Basic Engineering of the Gradual Reconstruction has been contracted to the Siemens AG. Implementation of safety measures are provided through contract with the consortium REKON (which consists of Siemens AG company and Nuclear Power Plants Research Institute Trnava) and other Czech, Russian and Slovak companies. The Gradual Reconstruction of Bohunice NPP V-1 will be finished in 2000. By implementation of the measures carried out during Gradual Reconstruction achievement of an internationally acceptable nuclear safety level will be reached. (author)

Analysis of core damage frequency: Nuclear power plant Dukovany, VVER/440 V-213 Unit 1, internal events. Volume 1: Main report

International Nuclear Information System (INIS)

Pugila, W.J.

1994-01-01

This report presents the final results from the Level 1 probabilistic safety assessment (PSA) for the Dukovany VVER/440 V-213 nuclear power plant, Unit 1. Section 1.1 describes the objectives of this study. Section 1.2 discusses the approach that was used for completing the Dukovany PSA. Section 1.3 summarizes the results of the PSA. Section 1.4 provides a comparison of the results of the Dukovany PSA with the results of other PSAs for different types of reactors worldwide. Section 1.5 summarizes the conclusions of the Dukovany PSA

Lessons learned from the seismic reevaluation of San Onofre Nuclear Generating Station, Unit 1

International Nuclear Information System (INIS)

Russell, M.J.; Shieh, L.C.; Tsai, N.C.; Cheng, T.M.

1987-01-01

A seismic reevaluation program was conducted for the San Onofre Nuclear Generating Station, Unit No. 1 (SONGS 1). SEP was created by the NRC to provide (1) an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The Systematic Evaluation Program (SEP) seismic review for SONGS 1 was exacerbated by the results of an evaluation of an existing capable fault near the site during the design review for Units 2 and 3, which resulted in a design ground acceleration of 0.67g. Southern California Edison Company (SCE), the licensee for SONGS 1, realized that a uniform application of existing seismic criteria and methods would not be feasible for the upgrading of SONGS 1 to such a high seismic requirement. Instead, SCE elected to supplement existing seismic criteria and analysis methods by developing criteria and methods closer to the state of the art in seismic evaluation techniques

Conformance to Regulatory Guide 1.97, Shearon Harris Nuclear Power Plant, Unit Nos. 1 and 2 (Docket Nos. 50-400 and 50-401)

International Nuclear Information System (INIS)

Stoffel, J.W.

1985-09-01

This EG and G Idaho, Inc. report reviews the submittals for Regulatory Guide 1.97, Revision 3, for Unit Nos. 1 and 2 of the Shearon Harris Nuclear Power Plant and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified. 7 refs

Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

Energy Technology Data Exchange (ETDEWEB)

Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

International Nuclear Information System (INIS)

Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

1995-07-01

The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

Science, society, and America's nuclear waste: Unit 3, The Nuclear Waste Policy Act

International Nuclear Information System (INIS)

1992-01-01

This teachers guide is unit 3, the nuclear waste policy act, in a four-unit secondary curriculum. It is intended to provide information about scientific and societal issues related to the management of spent nuclear fuel from generation of electricity at nuclear powerplants and high-level radioactive waste from US national defense activities. The curriculum, supporting classroom activities, and teaching materials present a brief discussion of energy and electricity generation, including that produced at nuclear power plants; information on sources, amounts, location, and characteristics of spent nuclear fuel and high-level radioactive waste; sources, types and effects of radiation; US policy for managing and disposing of spent nuclear fuel and high-level radioactive waste and what other countries are doing; and the components of the nuclear waste management system

Analysis for Human-related Events during the Overhaul

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Tae; Kim, Min Chull; Choi, Dong Won; Lee, Durk Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2011-10-15

The event frequency due to human error is decreasing among 20 operating Nuclear Power Plants (NPPs) excluding the NPP (Shin-Kori unit-1) in the commissioning stage since 2008. However, the events due to human error during an overhaul (O/H) occur annually (see Table I). An analysis for human-related events during the O/H was performed. Similar problems were identified for each event from the analysis and also, organizational and safety cultural factors were also identified

Development of PSI and ISI technique

International Nuclear Information System (INIS)

Chung, M.K.; Park, D.Y.; Choi, S.P.; Kim, H.J.; Moon, Y.S.; Shon, G.H.; Kim, T.S.

1983-01-01

This report describes the experimental results of the subjects selected from the PSI/ISI related problems which encountered by us in 1982. The main contents are 1) the characteristics of the typical ECT signals from the steam generator tubes of nuclear power plant and the results of ECT evaluation of Kori-1 steam generators, 2) the experimental result for the research for directional effects of ultrasonic transducers, 3) the basic experiment for the ultrasonic testing technique by immersion testing method, 4) how to write the scan plan of the mechanized ultrasonic testing for nuclear reactor. Attached appendix is a part of necessary materials for the scan plan of the mechanized ultrasonic testing for Kori-2 nuclear reactor. (Author)

Optimal sampling period of the digital control system for the nuclear power plant steam generator water level control

International Nuclear Information System (INIS)

Hur, Woo Sung; Seong, Poong Hyun

1995-01-01

A great effort has been made to improve the nuclear plant control system by use of digital technologies and a long term schedule for the control system upgrade has been prepared with an aim to implementation in the next generation nuclear plants. In case of digital control system, it is important to decide the sampling period for analysis and design of the system, because the performance and the stability of a digital control system depend on the value of the sampling period of the digital control system. There is, however, currently no systematic method used universally for determining the sampling period of the digital control system. Generally, a traditional way to select the sampling frequency is to use 20 to 30 times the bandwidth of the analog control system which has the same system configuration and parameters as the digital one. In this paper, a new method to select the sampling period is suggested which takes into account of the performance as well as the stability of the digital control system. By use of the Irving's model steam generator, the optimal sampling period of an assumptive digital control system for steam generator level control is estimated and is actually verified in the digital control simulation system for Kori-2 nuclear power plant steam generator level control. Consequently, we conclude the optimal sampling period of the digital control system for Kori-2 nuclear power plant steam generator level control is 1 second for all power ranges. 7 figs., 3 tabs., 8 refs. (Author)

Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 8

International Nuclear Information System (INIS)

1984-12-01

Supplement 8 (SSER 8) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 7

International Nuclear Information System (INIS)

1984-09-01

Supplement 7 (SSER 7) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety evaluation report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 6

International Nuclear Information System (INIS)

1984-07-01

Supplement 6 (SSER 6) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No. 1 (Docket No. 50-322). Supplement No. 9

International Nuclear Information System (INIS)

1985-12-01

Supplement 9 (SSER 9) to the Safety Evaluation Report on Long Island Lighting Company's application for a license to operate the Shoreham Nuclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses several items that have been reviewed by the staff since the previous supplement was issued

Draft environmental impact statement. River Bend Nuclear Power Station, Unit 1

International Nuclear Information System (INIS)

Anon.

1980-01-01

Federal financing of an undivided ownership interest of River Bend Nuclear Power Station Unit 1 on a 3293-acre site near St. Francisville, Louisiana is proposed in a supplement to the final environmental impact statement of September 1974. The facility would consist of a boiling-water reactor that would produce a maximum of 2894 megawatts (MW) of electrical power. A design level of 3015 MW of electric power could be realized at some time in the future. Exhaust steam would be cooled by mechanical cooling towers using makeup water obtained from and discharged to the Mississippi River. Power generated by the unit would be transmitted via three lines totaling 140 circuit miles traversing portions of the parishes of West Feliciana, East Feliciana, East Baton Rouge, West Baton Rouge, Pointe Coupee, and Iberville. The unit would help the applicant meet the power needs of rural electric consumers in the region, and the applicant would contribute significanlty to area tax base and employment rolls during the life of the unit. Construction related activities would disturb 700 forested acres on the site and 1156 acres along the transmission routes. Of the 60 cubic feet per second (cfs) taken from the river, 48 cfs would evaporate during the cooling process and 12 cfs would return to the river with dissolved solids concentrations increased by 500%. The terrace aquifer would be dewatered for 16 months in order to lower the water table at the building site, and Grants Bayou would be transformed from a lentic to a lotic habitat during this period. Fogging and icing due to evaporation and drift from the cooling towers would increase slightly. During the construction period, farming, hunting, and fishing on the site would be suspended, and the social infractructure would be stressed due to the influx of a maximum of 2200 workers

Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2

International Nuclear Information System (INIS)

1995-09-01

This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985). Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992). Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April 1993), Supplement No. 12 (October 1993). Supplement No. 13 (April 1994), Supplement No. 14 (December 1994), and Supplement No. 15 (June 1995) issued by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos, 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER

IAEA Completes Nuclear Security Review Mission in United States

International Nuclear Information System (INIS)

2013-01-01

Full text: A team of nuclear security experts led by the International Atomic Energy Agency (IAEA) today completed a mission to review nuclear security practices of civil nuclear facilities licensed by the United States Nuclear Regulatory Commission (NRC). Conducted at the U.S. Government's request, the two-week International Physical Protection Advisory Service (IPPAS) mission reviewed the United States' nuclear security-related legislative and regulatory framework. As part of this work, the IPPAS team, led by John O'Dacre of Canada and comprising nine experts from eight IAEA Member States, met with NRC officials and reviewed the physical protection systems at the Center for Neutron Research (NCNR) at the National Institute of Standards and Technology. The IPPAS team concluded that nuclear security within the U.S. civil nuclear sector is robust and sustainable and has been significantly enhanced in recent years. The team identified a number of good practices in the nation's nuclear security regime and at the NCNR. The IPPAS team also made a recommendation and some suggestions for the continuing improvement of nuclear security overall. The mission in the United States was the 60th IPPAS mission organized by the IAEA. 'Independent international peer reviews such as IAEA IPPAS missions are increasingly being recognized for their value as a key component for exchanges of views and advice on nuclear security measures', said Khammar Mrabit, Director of the IAEA Office of Nuclear Security. 'The good practices identified during this mission will contribute to the continuous improvements of nuclear security in other Member States'. The IPPAS team provided a draft report to the NRC and will submit a final report soon. Because it contains security-related information about a specific nuclear site, IPPAS reports are not made public. 'The IPPAS programme gives us a chance to learn from the experience and perspective of our international partners', said NRC Chairman Allison M

Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 (Docket Nos. 50-416 and 50-417). Supplement No. 5

International Nuclear Information System (INIS)

1984-08-01

Supplement 5 to the Safety Evaluation Report for Mississippi Power and Light Company, et al., joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status on the resolution of those issues that require further evaluation before authorizing operation of Unit 1 above 5% of rated power

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Sohrabi, M.; Ghasemi, M.; Amrollahi, R.; Khamooshi, C.; Parsouzi, Z. [Amirkabir University of Technology, Health Physics and Dosimetry Research Laboratory, Department of Physics, Tehran (Iran, Islamic Republic of)

2013-05-15

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well

Experiences and plot application of maintenance rule for nuclear power plant in Korea

International Nuclear Information System (INIS)

Choi, Kwang Hee; Seo, Mi Ro; Kim, Myung Ki; Yoo, Sung Soo

2007-01-01

In order to enhance the effectiveness of maintenance, Korea lunched two pilot project to implement the Maintenance Rule(MR) program similar to that of USA. One for the pilot program is Ulchin units 3 and 4, Korean standardized nuclear power(KSNP) Plant, which was started in October 2003 and finished in September 2006. The other program is for Kori units 3 and 4, Westinghouse type plants, which was stared in April 2004 and completed in September 2006. MR processes consists of 'scope determination', 'safety significance determination', 'performance criteria development', 'performance monitoring' and 'disposition process between enforcement monitoring and routine monitoring'. At each process, we gave a expert panel to review output of each process and make a justification on the output. these expert panels meeting made much of technical discussions and decisions of major products on each process. In addition, Maintenance effectiveness and Target observation system(MENTOS) was developed to manage MR implementation, of which data is automatically imported from Enterprise Resource Program(ERP), and outputs are exported to ERP and the final results are shown in ERP screen. In this paper, major results and experiences gained from each application process at Ulchin 3,4 nuclear power plants were introduced

United States - Japanese nuclear relations: implications for the pacific region

International Nuclear Information System (INIS)

Suttmeier, R.P.

1980-01-01

The initiation of a new approach to non-proliferation policy by the United States in 1977 was the most upsetting development in the history of US-Japanese nuclear relations. The policy has seemingly altered Japanese views of its own nuclear future very little, yet it has altered the tone of the nuclear relationship with the United States. Recent Japanese nuclear power developments, both technical and administrative, are outlined and the prospects for the future of the United States-Japanese nuclear relationship and for Pacific regional cooperation assessed. Issues of importance in the relationship include reprocessing - enrichment, plutonium management, spent fuel and waste management and uranium supplies

Safety-evaluation report related to the operation of Diablo Canyon Nuclear Power Plants, Units 1 and 2. Docket Nos. 50-275 and 50-323. Supplement No. 18

International Nuclear Information System (INIS)

1983-08-01

Supplement 18 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the verification effort for Diablo Canyon Unit 1 that was performed between November 1981 and the present in response to Commission Order CLI-81-30 and an NRC letter to the licensee

Preoperation of Hamaoka Nuclear Power Station Unit No. 4

International Nuclear Information System (INIS)

Fukuyo, Tadashi; Kurata, Satoshi

1994-01-01

Chubu Electric Power Co. finished preoperation of Hamaoka Nuclear Power Station Unit No. 4 in September, 1993. Although unit 4 has the same reactor design as unit 3, its rated electrical output (1,137MW) is 37MW more than that of unit 3. This increase was achieved mainly by adopting a Moisture Separater Heater in the turbine system. We started preoperation of unit 4 in November 1992 and performed various tests at electrical outputs of 20%, 50%, 75%, and 100%. We finished preoperation without any scram or other major problems and obtained satisfactory results for the functions and performance of the plant. This paper describes the major results of unit 4 preoperation. (author)

The nuclear energy in the United Kingdom

International Nuclear Information System (INIS)

2006-02-01

With challenges like the climatic change, the hydrocarbons prices increase and the energy supply security, the nuclear park is becoming a decisive and an urgent question in the United Kingdom. The author proposes an historical aspect of the nuclear energy in UK, the actors of the today nuclear industry and the technologies used in 2006, the radioactive wastes management, the programs of the future and the british opinion on the nuclear. (A.L.B.)

Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

International Nuclear Information System (INIS)

Brown, R.; Cobanoglu, M.M.

1995-01-01

Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

Optimal inspection and replacement periods of the safety system in Wolsung Nuclear Power Plant Unit 1 with an optimized cost perspective

International Nuclear Information System (INIS)

Jinil Mok; Poong Hyun Seong

1996-01-01

In this work, a model for determining the optimal inspection and replacement periods of the safety system in Wolsung Nuclear Power Plant Unit 1 is developed, which is to minimize economic loss caused by inadvertent trip and the system failure. This model uses cost benefit analysis method and the part for optimal inspection period considers the human error. The model is based on three factors as follows: (i) The cumulative failure distribution function of the safety system, (ii) The probability that the safety system does not operate due to failure of the system or human error when the safety system is needed at an emergency condition and (iii) The average probability that the reactor is tripped due to the failure of system components or human error. The model then is applied to evaluate the safety system in Wolsung Nuclear Power Plant Unit 1. The optimal replacement periods which are calculated with proposed model differ from those used in Wolsung NPP Unit 1 by about a few days or months, whereas the optimal inspection periods are in about the same range. (author)

Seismic structural fragility investigation for the San Onofre Nuclear Generating Station, Unit 1 (Project I); SONGS-1 AFWS Project

International Nuclear Information System (INIS)

Wesley, D.A.; Hashimoto, P.S.

1982-04-01

An evaluation of the seismic capacities of several of the San Onofre Nuclear Generating Station, Unit 1 (SONGS-1) structures was conducted to determine input to the overall probabilistic methodology developed by Lawrence Livermore National Laboratory. Seismic structural fragilities to be used as input consist of median seismic capacities and their variabilities due to randomness and uncertainty. Potential failure modes were identified for each of the SONGS-1 structures included in this study by establishing the seismic load-paths and comparing expected load distributions to available capacities for the elements of each load-path. Particular attention was given to possible weak links and details. The more likely failure modes were screened for more detailed investigation

Pickering Unit 1 chemical cleaning

International Nuclear Information System (INIS)

Smee, J.L.; Fiola, R.J.; Brennenstuhl, K.R.; Zerkee, D.D.; Daniel, C.M.

1995-01-01

The secondary sides of all 12 boilers at Pickering Unit 1 were chemically cleaned in 1994 by the team of Ontario Hydro, B and W International (Cambridge, Ontario) and B and W Nuclear Technologies (Lynchburg, Virginia). A multi-step EPRI/SGOG process was employed in a similar manner to previous clearings at Units 5 and 6 in 1992 and 1993, respectively. A major innovation with the Unit 1 cleaning was the incorporation of a crevice cleaning step, the first time this had been done on Ontario Hydro plants. In addition, six boilers were cleaned in parallel compared to three at a time in previous Pickering cleanings. This significantly reduced cleaning time. A total of 6,770 kg of sludge was removed through direct chemical dissolution. It consisted of 66% iron/nickel oxides and 28% copper metal. A total of 1,600,000 L (420,000 US gallons) of liquid waste was produced. It was processed through the spent solvent treatment facility located at the Bruce Nuclear Power Development site. Visual inspection performed after the cleaning indicated that the crevices between the boiler tubes and the tube support structure were completely clear of deposit and the general condition of the tubing and lattice bars appeared to be in 'as new' condition. (author)

Technical Specifications, Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. 50-400). Appendix ''A'' to License No. NPF-53

International Nuclear Information System (INIS)

1986-10-01

This report presents specifications for the Shearon Harris Nuclear Power Plant Unit No. 1 concerning: safety limits and limiting safety system settings; limiting conditions for operation and surveillance requirements; design features; and administrative controls

Dresden Nuclear Power Station, Unit No. 1: Primary cooling system chemical decontamination: Draft environmental statement (Docket No. 50-10)

International Nuclear Information System (INIS)

1980-05-01

The staff has considered the environmental impact and economic costs of the proposed primary cooling system chemical decontamination at Dresden Nuclear Power Station, Unit 1. The staff has focused this statement on the occupational radiation exposure associated with the proposed Unit 1 decontamination program, on alternatives to chemical decontamination, and on the environmental impact of the disposal of the solid radioactive waste generated by this decontamination. The staff has concluded that the proposed decontamination will not significantly affect the quality of the human environment. Furthermore, any impacts from the decontamination program are outweighed by its benefits. 2 figs., 7 tabs

Science, Society, and America's Nuclear Waste: The Nuclear Waste Policy Act, Unit 3. Teacher Guide. Second Edition.

Science.gov (United States)

Department of Energy, Washington, DC. Office of Civilian Radioactive Waste Management, Washington, DC.

This guide is Unit 3 of the four-part series, Science, Society, and America's Nuclear Waste, produced by the U.S. Department of Energy's Office of Civilian Radioactive Waste Management. The goal of this unit is to identify the key elements of the United States' nuclear waste dilemma and introduce the Nuclear Waste Policy Act and the role of the…

Heat balance calculation and feasibility analysis for initial startup of Fuqing nuclear turbine unit with non-nuclear steam

International Nuclear Information System (INIS)

He Liu; Xiao Bo; Song Yumeng

2014-01-01

Non-nuclear steam run up compared with nuclear steam run up, can verify the design, manufacture, installation quality of the unit, at the same time shorten the follow-up duration of the entire group ready to start debugging time. In this paper, starting from the first law of thermodynamics, Analyzed Heat balance Calculation and Feasibility analysis for Initial startup of Fuqing nuclear Turbine unit with Non-nuclear steam, By the above calculation, to the system requirements and device status on the basis of technical specifications, confirmed the feasibility of Non-nuclear steam running up in theory. After the implementation of the Non-nuclear turn of Fuqing unit, confirmed the results fit with the actual process. In summary, the Initial startup of Fuqing turbine unit with Non-nuclear steam is feasible. (authors)

Conformance to Regulatory Guide 1.97 Browns Ferry Nuclear Plant, Unit Nos. 1, 2 and 3 (Docket Nos. 50-259, 50-260, and 50-296)

International Nuclear Information System (INIS)

Udy, A.C.

1985-07-01

This EG and G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97, Revision 3, for Unit Nos. 1, 2 and 3 of the Browns Ferry Nuclear Plant and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified

Safety assessment of Olkiluoto NPP units 1 and 2. Decision of the Radiation and Nuclear Safety Authority regarding the periodic safety review of the Olkiluoto NPP

International Nuclear Information System (INIS)

2010-02-01

In this safety assessment the Radiation and Nuclear Safety Authority (STUK) has evaluated the safety of the Olkiluoto Nuclear Power Plant units 1 and 2 in connection with the periodic safety review. This safety assessment provides a summary of the reviews, inspections and continuous oversight carried out by STUK. The issues addressed in the assessment and the related evaluation criteria are set forth in the nuclear energy and radiation safety legislation and the regulations issued thereunder. The provisions of the Nuclear Energy Act concerning the safe use of nuclear energy, security and emergency preparedness arrangements, and waste management are specified in more detail in the Government Decrees and Regulatory Guides issued by STUK. Based on the assessment, STUK consideres that the Olkiluoto Nuclear Power Plant units 1 and 2 meet the set safety requirements for operational nuclear power plants, the emergency preparedness arrangements are sufficient and the necessary control to prevent the proliferation of nuclear weapons has been appropriately arranged. The physical protection of the Olkiluoto nuclear power plant is not yet completely in compliance with the requirements of Government Decree 734/2008, which came into force in December 2008. Further requirements concerning this issue based also on the principle of continuous improvement were included in the decision relating to the periodic safety review. The safety of the Olkiluoto nuclear power plant was assessed in compliance with the Government Decree on the Safety of Nuclear Power Plants (733/2008), which came into force in 2008. The decree notes that existing nuclear power plants need not meet all the requirements set out for new plants. Most of the design bases pertaining to the Olkiluoto 1 and 2 nuclear power plant units were set in the 1970s. Substantial modernisations have been carried out at the Olkiluoto 1 and 2 nuclear power plant units since their commissioning to improve safety. This is in line with

78 FR 29158 - In the Matter of Zion Solutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Order Approving...

Science.gov (United States)

2013-05-17

... and DPR-48] In the Matter of Zion Solutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Order... formed for the purpose of acquiring ES, Inc. and is held by certain investment fund entities organized by... Environmental Management Programs, in writing, of such receipt no later than one (1) business day prior to the...

Operating performance of LWR nuclear generating units

International Nuclear Information System (INIS)

Pia, S.

1984-01-01

This work aims at reviewing, on the basis of historical data, the operational problem areas which explain the degree of availability and productivity achieved up to now by nuclear power plants in commercial operation in the world. The operating performance data of nuclear power plants area analysed with respect to plant type, size and other significant reference parameters and they are evaluated also by comparison with fossil generating unit data. Major performance indices data are presented for both nuclear and fossil units type and distribution of outage causes. Unplanned full outages caused by nuclear power plant equipment and components failure are particulary emphasized. The trend for unplanned full outages due to the failure of components shows decreasing numerical values in 1981 with respect to the previous years. But this result should be weighed with the increasing plant unavailability hours needed for maintenance and repair action (chiefly preventive maintenance on critical components). This means that the number and downtime of forced outage must be drastically reduced for economic reasons (production losses and problems associated with the unavailable unit unplanned replacement) as well as for plant safe and reliable operation (sudden unavailability of key components and frequency of transients associated with plant shutdown and routine startup operation)

Law on the Decommissioning of unit 1 at the state enterprise of the Republic of Lithuania Ignalina Nuclear Power Plant

International Nuclear Information System (INIS)

2000-01-01

This law regulates the legal principles for the decommissioning of unit 1 at the Ignalina Nuclear Power Plant. The main deadlines for the government in the preparation for the decommissioning are set in the law. All preparatory works should be finished before the year 2005

Nuclear instrument maintenance and technical training in Nuclear Energy Unit

International Nuclear Information System (INIS)

Mohamad Nasir Abdul Wahid

1987-01-01

Instrument maintenance service is a necessity in a Nuclear Research Institute, such as the Nuclear Energy Unit (NEU) to ensure the smooth running of our research activities. However, realising that maintenance back-up service for either nuclear or other scientific equipment is a major problem in developing countries such as Malaysia, NEU has set up an Instrumentation and Control Department to assist in rectifying the maintenance problem. Beside supporting in house activities in NEU, the Instrumentation and Control Department (I and C) is also geared into providing services to other organisations in Malaysia. This paper will briefly outline the activities of NEU in nuclear instrument maintenance as well as in technical training. (author)

Kozloduy nuclear power plant. Units 1-4. Status of safety assessment activities. Rev. 2

International Nuclear Information System (INIS)

1999-01-01

This paper presents the results of the status of safety assessment activities carried out by the Kozloduy Nuclear Power Plant (KNPP) in order to evaluate the current status of the safety of its reactor units 1-4. The steam supply system of this units is based of the reactor WWER-440/ B-230, which is a PWR of Russian design developed according to the safety standards in force in USSR in late 60-s. Now a days 10 reactor units of this type are in operation in four NPPs. Despite of efforts of the different plants to implement safety improvements measures during first 10-15 years of operation of this type of reactor its major safety problems were not eliminated and were a subject of international concern. The systematic evaluation of the deficiencies of the original design of this type of reactors have been initiated by IAEA in the beginning of 1990 and brought to developing a comprehensive list of safety problems which required urgent implementation of safety measures in all plants. To solve this problems in 1991 KNPP initiated implementation of so called 'short term' safety improvement program, developed with the help of WANO under agreement with Bulgarian Nuclear Safety Authority (BNSA) and consortium RISKAUDIT. The program was based on a stage approach and was foreseen to be implemented by tree stages in very tight time schedule in order to achieve significant and rapid improvements of the level of safety in operation of the units. The Short Tenn Program was implemented between the years 1991 and 1997 thanks of the strong safety commitment of NEK and KNPP staff and the broad international cooperation and financial support. Important part of resources were supplied under PHARE program of CEC, EBRD grant agreement and EDF support. The plant current safety level analysis has been performed using IAEA analytical methodology according to 50-SG-O12 standard 'Periodic safety review of operational nuclear power plants'. The approach and criteria for acceptable safety level

The future of nuclear power in the United Kingdom

International Nuclear Information System (INIS)

Jones, P.M.S.

1990-01-01

The arguments that have been put forward in the United Kingdom to justify the contention by its opponents that nuclear power should not be seen as a safe, economic, strategically desirable and environmentally attractive energy source are examined. Counter arguments are presented to support the belief that these are wholly wrong. In the short to medium term, however, economic and political considerations suggest that the prospects for nuclear power in the United Kingdom are not optimistic. The long term evolution of nuclear power is uncertain but it is possible that the security of energy supply, comparative economics and environmental attractions associated with nuclear power will become more apparent and work in its favour eventually. The conviction is expressed that, by the middle of the next century, the United Kingdom will be reaping the benefits of the fast reactor on a significant scale. (UK)

Safety Evaluation Report related to the operation of Cartawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414). Supplement No. 5

International Nuclear Information System (INIS)

1986-02-01

This report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, Saluda River Electric Cooperative, Inc., and Piedmont Municipal Power Agency, as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for initial criticality and power ascension to full-power operation for Unit 2

Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

International Nuclear Information System (INIS)

Kobayashi, Masahide; Narabayashi, Tadashi; Tsuji, Masashi; Chiba, Go; Nagata, Yasunori; Shimoe, Tomohiro

2015-01-01

As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained. (author)

Role and services of the Nuclear Energy Unit

International Nuclear Information System (INIS)

Ahmad Tajuddin bin Ali

1985-01-01

This article contains the research and development activities planned to be undertaken, and the services to be provided by the Nuclear Energy Unit. The fields of research have been selected to aid the development in the use of Nuclear Science and Technology, whilst the proposed services are geared towards widening the use to this technology in socio-economic development. Objectives in the research fields have been identified following a number of consultations with IAEA, foreign and local experts, and involved careful study of local requirements. The services provided will cater for all, especially institutions involved in nuclear technology both from the public and private sectors. This effort is in line with the role of the Unit as the national promoter and coordinator in the utilization of nuclear technology in the country

Peer review of the Barselina Level 1 probabilistic safety assessment of the Ignalina Nuclear Power Plant, Unit 2

International Nuclear Information System (INIS)

McKay, S.L.; Coles, G.A.

1995-01-01

The Barselina Project is a Swedish-funded, cooperative effort among Lithuania, Russia and Sweden to transfer Western probabilistic safety assessment (PSA) methodology to the designers/operators of Ignalina Nuclear Power Plant (INPP). The overall goal is to use the PSA as a tool for assessing plant operational safety. The INPP is a two-unit, Former Soviet Union-designed nuclear facility located in Lithuania. The results of this PSA will ultimately be used to identify plant-specific improvements in system design and the conduct of facility operations, allowing improved operational safety. Pacific Northwest Laboratory (PNL) was asked to perform an independent expert peer review of the Barselina PSA. This report documents the findings of this review. This review, financed with nuclear safety assistance funds through the US Agency for International Development (USAID) and the US Department of Energy (DOE), satisfies Task II of the PNL peer review of the Barselina project. The objective is to provide an independent, in-proce ss examination of the Barselina Level 1 PSA of Ignalina Nuclear Power Plant, Unit 2. The review consisted of an investigation of the project documentation, interviews, and extensive discussions with the PSA staff during critical stages of the project. PNL assessed the readability, completeness, consistency, validity, and applicability of the PSA. The major aspects explored were its purpose, major assumptions, analysis/modeling, results, and interpretation. It was not within the scope of this review to perform plant walkdowns or to review material other than the PSA documentation

Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414. Suppl. 1

International Nuclear Information System (INIS)

1983-04-01

This reort supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report, and discusses the recommendations of the Advisory Committee on Reactor Safeguards in its report dated March 15, 1983

Impact assessment on the human exposure to environmental radiation

International Nuclear Information System (INIS)

Kim, Y.J.; Lee, K.S.; Kim, J.B.; Chun, K.J.; Kim, S.R.; Chung, K.H.

1983-01-01

1. The Terrestrial Ecosystem in the Vicinity of Kori Nuclear Power Plant. In order to evaluate the effects of radiation or radionuclides released from the nuclear power plants on human population, field surveys on the terrestrial ecosystem such as the fauna and flora in the vicinity within 15 km from the Kori nuclear power plant were carried out. 2. Level of radionuclides, Metabolism and Chromosome Aberration. The present study was undertaken with the purpose for evaluating 1) the level of Sr-90 in the Korean vertebra, 2) the metabolism of Sr-90 in the rats, 3) the levels of Sr-90 and Cs-137 in the pine needles (Pinus densiflora), 4) the level of Sr-90 in the frogs (Rana nigromaculata) and 5) the radiation effects by internal Sr-90 on the bone marrow. (Author)

Nuclear safety philosophy in the United Kingdom

International Nuclear Information System (INIS)

Anthony, R.D.

1986-01-01

Development of the United Kingdom (UK) nuclear safety philosophy is described in the context of the UK nuclear power program since 1959 and of its legislative framework. Basic to the philosophy is that the licensee is wholly responsible for nuclear safety. The licensing process and safety assessment principles used by the Nuclear Installations Inspectorate are discussed, and examples from the assessment of the proposed UK pressurized-water reactor are used to illustrate how the approach works in practice. The UK siting policy and regulatory developments since 1979 are also discussed. Recent, current, and future issues of interest to the regulatory authority are described against the development nuclear scene in the UK

Project for construction of a new nuclear unit in Armenia

International Nuclear Information System (INIS)

2010-01-01

WorleyParsons Project Description Scope of work: Management Services for Implementation of the Project on Construction of New Nuclear Unit(s) in the Republic of Armenia. Client: Ministry of Energy and Natural Resources (MENR), Republic of Armenia •Contract signed on 29 of May 2009 •Project divided into 4 Phases. Addendum 1 covers phase I and II signed on 27 of July. Contract entered into force 7 of August

Draft Environmental Impact Statement on a proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel. Volume 1

International Nuclear Information System (INIS)

1995-03-01

The United States Department of Energy and United States Department of State are jointly proposing to adopt a policy to manage spent nuclear fuel from foreign research reactors. Only spent nuclear fuel containing uranium enriched in the United States would be covered by the proposed policy. The purpose of the proposed policy is to promote U.S. nuclear weapons nonproliferation policy objectives, specifically by seeking to reduce highly-enriched uranium from civilian commerce. Environmental effects and policy considerations of three Management Alternative approaches for implementation of the proposed policy are assessed. The three Management Alternatives analyzed are: (1) acceptance and management of the spent nuclear fuel by the Department of Energy in the United States, (2) management of the spent nuclear fuel at one or more foreign facilities (under conditions that satisfy United States nuclear weapons nonproliferation policy objectives), and (3) a combination of components of Management Alternatives 1 and 2 (Hybrid Alternative). A No Action Alternative is also analyzed. For each Management Alternative, there are a number of alternatives for its implementation. For Management Alternative 1, this document addresses the environmental effects of various implementation alternatives such as varied policy durations, management of various quantities of spent nuclear fuel, and differing financing arrangements. Environmental impacts at various potential ports of entry, along truck and rail transportation routes, at candidate management sites, and for alternate storage technologies are also examined. For Management Alternative 2, this document addresses two subalternatives: (1) assisting foreign nations with storage; and (2) assisting foreign nations with reprocessing of the spent nuclear fuel. Management Alternative 3 analyzes a hybrid alternative. This document is Vol. 1 of 2 plus summary volume

Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 26

International Nuclear Information System (INIS)

1984-07-01

Supplement 26 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. The supplement reports on the status of the staff's investigation, inspection and evaluation of those allegations or concerns that have been identified to the NRC as of July 8, 1984. The report specifically addresses those allegations which the staff determined must be satisfactorily resolved prior to full power operation of Diablo Canyon Unit 1

The operation of post-irradiation examination facility

International Nuclear Information System (INIS)

Kim, Eun Ka; Min, Duk Ki; Lee, Young Kil

1994-12-01

The operation of post-irradiation examination facility was performed as follow. HVAC and pool water treatment system were continuously operated, and radiation monitoring in PIE facility has been carried out to maintain the facility safely. Inspection of the fuel assembly (F02) transported from Kori Unit 1 was performed in pool, and fuel rods extracted from the fuel assembly (J44) of Kori Unit 2 NPP were examined in hot cell. A part of deteriorated pipe line of drinking water was exchanged for stainless steel pipe to prevent leaking accidents. Halon gas system was also installed in the exhausting blower room for fire fighting. And IAEA inspection camera for safeguard of nuclear materials was fixed at the wall in pool area. Radiation monitoring system were improved to display the area radioactive value at CRT monitor in health physics control room. And automatic check system for battery and emergency diesel generator was developed to measure the voltage and current of them. The performance test of oxide thickness measuring device installed in hot cell for irradiated fuel rod and improvement of the device were performed, and good measuring results using standard sample were obtained. The safeguard inspection of nuclear materials and operation inspection of the facility were carried out through the annual operation inspection, quarterly IAEA inspection and quality assurance auditing. 26 tabs., 43 figs., 14 refs. (Author) .new

Report of the international fire safety mission to Temelin, unit 1 nuclear power plant Czech Republic 4 to 14 February 1996

International Nuclear Information System (INIS)

1996-01-01

This report presents the results of an IAEA Fire Safety Mission conducted within the scope of Technical Co-operation Project CZR/9/005 to assess the licensing process, design, analysis and operational management of the Temelin Nuclear Power Plant with regards to fire safety of the plant. The Temelin Nuclear Power Plant currently has two units under construction. Each unit is equipped with a pressurized water reactor of the WWER design with a net electrical output of about MWe. The plant has already made significant upgrading in fire protection from the original design. The Team's evaluation is based on the IAEA Safety Series No. 50-SG-D2 (Rev.1), Fire Protection in Nuclear Power Plants, and other fire protection guidelines currently produced by the IAEA. The evaluation, conclusions and recommendations presented in this report reflect the views of the Fire Safety Mission experts. The recommendations are provided for consideration by the responsible authorities in the Czech Republic towards enhancing fire safety at the Temelin plant

Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391

International Nuclear Information System (INIS)

1992-01-01

Supplement No. 8 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation of (1) additional information submitted by the applicant since Supplement No. 7 was issued, and (2) matters that the staff had under review when Supplement No. 7 was issued

78 FR 47011 - Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

Science.gov (United States)

2013-08-02

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0195] Software Unit Testing for Digital Computer Software... revised regulatory guide (RG), revision 1 of RG 1.171, ``Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.'' This RG endorses American National Standards...

Safety Evaluation Report related to the operation of Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414). Supplement No. 6

International Nuclear Information System (INIS)

1986-05-01

This report supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, Saluda River Electric Cooperative, Inc., and Piedmont Municipal Power Agency, as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 miles) north of Rock Hill and adjacent to Lake Wylie. This supplement provides additional information supporting the license for operation above 5% power and power ascension to full-power operation for Unit 2

Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 17

International Nuclear Information System (INIS)

Tam, P.S.

1995-10-01

This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), 1991), Supplement No. 8 (January 1992), Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April.1993), Supplement No. 12 (October 1993), Supplement No. 13 (April 1994), Supplement No. 14 (December 1994), Supplement No. 15 (June 1995), and Supplement No. 16 (September 1995) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50--390 and 50--391). The facility is located in Rhea county, Tennessee, near the Watts Bar Dam on the Tennessee River. In this supplement, NRC examines the significant problems of construction quality and quality assurance effectiveness that led TVA to withdraw its certification in 1985 that Watts Bar Unit I was ready to load fuel. Also discussed are the extensive corrective actions performed by TVA according to its nuclear performance plans and other supplemental programs, and NRC's extensive oversight to determine whether the Watts Bar Unit 1 construction quality and TVA's operational readiness and quality assurance effectiveness are adequate for a low-power operating license to be issued. SSER 17 does not address Watts Bar Unit 2, except for the systems which are necessary to support Unit 1 operation

Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 4

International Nuclear Information System (INIS)

1986-10-01

This report, Supplement No. 4 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicants) for a license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1, 2, and 3

Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 2

International Nuclear Information System (INIS)

1985-06-01

This report, Supplement No. 2 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicants) for a license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement No. 1

Safety Evaluation Report related to the operation of Shearon Harris Nuclear Power Plant, Unit No. 1 (Docket No. STN 50-400). Supplement No. 3

International Nuclear Information System (INIS)

1986-05-01

This report, Supplement No. 3 to the Safety Evaluation Report for the application filed by the Carolina Power and Light Company and North Carolina Eastern Municipal Power Agency (the applicants) for a license to operate the Shearon Harris Nuclear Power Plant Unit 1 (Docket No. 50-400), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and Supplement Nos. 1 and 2

Failure analysis of retired steam generator tubings

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

2005-04-15

Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

Nuclear fuel element recovery using PEDSCO RMI Unit

International Nuclear Information System (INIS)

Martin, D.G.; Pedersen, B.V.

1984-01-01

In September 1982, a PEDSCO Remote Mobile Investigation Unit was used to recover damaged irradiated fuel elements from a fueling machine and trolley deck at Bruce Nuclear Generating Station 'A'. This Canadian-made remote controlled vehicle was originally designed for explosive ordinance disposal by law enforcement agencies. This paper describes its adaptation to nuclear service and its first mission, within a nuclear facility

Cernavoda NPP Unit 1 - a plant of several generations

International Nuclear Information System (INIS)

Rotaru, I.; Metes, M.; Anghelescu, M.S.

2001-01-01

The paper reflects some key aspects related to the shift of generations during the project's development, including the present stage. Further, the place of Cernavoda NPP Unit 1 in the Romanian power sector and among other nuclear stations in the world is presented. The operational performances achieved 'in service' up to the end of 1999, with reference to the performance indicators for electrical energy production, nuclear safety, radiation protection, radioactive wastes and nuclear fuel are illustrated. For all of these items, comparisons are performed with similar indicators reported by other worldwide nuclear power plants, in order to assess our results. Finally, some comments about Cernavoda NPP Unit 2 project status and need to completion and commissioning it are included. (authors)

Pressure tube replacement in Pickering NGS A units 1 and 2

International Nuclear Information System (INIS)

Irvine, H.S.; Bennett, E.J.; Talbot, K.H.

1986-10-01

Being able to technically and economically replace the most radioactive components (excluding the nuclear fuel) in operating reactors will help to ensure the ongoing acceptance of nuclear power as a viable energy source for the future. Ontario Hydro is well along the path to meeting the above objective for its CANDU-PHW reactors. Following the failure of a Zircaloy-II pressure tube in unit 2 of Pickering NGS A in August, 1983, Ontario Hydro has embarked on a program to replace all Zircaloy-II pressure tubes in units 1 and 2 at Pickering. This program integrates the in-house research, design, construction, and operating skills of a large utility (Ontario Hydro) with the skills of a national nuclear organization (Atomic Energy of Canada Limited) and the private engineering sector of the Canadian nuclear industry. The paper describes the background to the pressure tube failure in Pickering unit 2 and to the efforts incurred in understanding the failure mechanism and how similar failures are not expected for the zirconium-niobium pressure tube material used in all other large CANDU-PHW units after units 1 and 2 of Pickering NGS A. The tooling developed for the pressure tube replacement program is described as well as the organization to undertake the program in an operating nuclear station. The retubing of units 1 and 2 at Pickering NGS A is nearing a successful completion and shows the benefits of being able to integrate the various skills required for this success. Pressure tube replacement in a CANDU-PHW reactor is equivalent to replacement of the reactor vessel in a LWR. The fact that this replacement can be done economically and with acceptable radiation dose to workers augurs well for the continued viability of the use of nuclear energy for the benefit of mankind. (author)

United Nuclear Industries, Inc. reactor and fuel production facilities 1975 environmental release report

International Nuclear Information System (INIS)

Cucchiara, A.L.

1976-01-01

During calendar year 1975, an estimated total of 3,000,000 pounds of waste materials and approximately 150 curies of radionuclides were discharged to the environs in liquid effluent streams emanating from United Nuclear Industries, Inc., operated facilities. During the same period, approximately 1,700,000 pounds of reported waste materials, including 34,000 curies of reported radionuclides, were discharged to the atmosphere from United Nuclear Industries, Inc., operated facilities. Superscript numbers reference explanatory notes contained at the end of the report

The Iranian nuclear deal. Caught between centrifuge and separative working unit?

Energy Technology Data Exchange (ETDEWEB)

Ghoshal, Debalina

2017-11-15

In November 2013, a historic nuclear deal, the Joint Plan of Action (JPA) was struck between Iran and the P5+1 Countries, United States, Russia, China, Britain, France and Germany. Under this deal, Iran agreed to restrict its nuclear program that would close the doors for Iran to develop nuclear weapons. This indeed was historic as for over a decade, ever since the Iranian nuclear program attracted international attention and criticism, several arrangements were made to curtail the Iranian nuclear program, but none reached the desired success. In exchange for abiding by the JPA, Iran was promised partial relief of sanctions. The paper tracks the development and identifies the scope of Iran's centrifuge program.

Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 25

International Nuclear Information System (INIS)

1984-07-01

Supplement 25 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Unit 1 and Unit 2 (Docket Nos. 50-275 and 50-323) has been prepared by the Office of Nuclear Reactor Regulation (NRR) of the US Nuclear Regulatory Commission. This supplement reports on the staff's inspection and evaluation efforts on the matter of piping and piping supports as related to the seven technical license conditions in an Order Modifying License issued by NRR on April 18, 1984

User's guide for PRISM (Plant Risk Status Information Management System) Arkansas Nuclear One-Unit 1: Volume 1, Program for inspectors

International Nuclear Information System (INIS)

Campbell, D.J.; Guthrie, V.H.; Kirchner, J.R.; Kirkman, J.Q.; Paula, H.M.; Ellison, B.C.; Dycus, F.M.; Farquharson, J.A.; Flanagan, G.F.

1988-03-01

This user's guide is a two-volume document designed to teach NRC inspectors and NRC regulators how to access probabilistic risk assessment information from the two Plant Risk Status Information Management System (PRISIM) programs developed for Arkansas Nuclear One -- Unit One (ANO-1). This document, Volume 1, describes how the PRA information available in Version 1.0 of PRISIM is useful for planning inspections. Using PRISIM, inspectors can quickly access PRA information and use that information to update risk analysis results, reflecting a plant's status at any particular time. Both volumes are stand-alone documents, and each volume presents several sample computer sessions designed to lead the user through a variety of PRISIM applications used to obtain PRA-related information for monitoring and controlling plant risk

New United States policies regarding international nuclear cooperation

International Nuclear Information System (INIS)

Marshall, H.R. Jr.

1981-10-01

This paper discusses the United States policy on international nuclear power development in the light of the priorities established by President Reagan in the guidelines for his Administration's nuclear co-operation policy. The aim is to establish a framework allowing for co-operation in peaceful nuclear development while remaining committed to the objective of preventing the further spread of nuclear weapons, in particular by supporting the Non-Proliferation Treaty, the IAEA Safeguards System and the Tlatelolco Treaty (NEA) [fr

Development of Soil Derived Concentration Guidance Levels for Decommissioning at Overseas Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Sohn, Wook; Yoon, Suk Bon; Kim, Jeongju [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In Korea, the criteria are expected to be given in terms of dose as in US and Spain. However, since dose cannot be measured, corresponding measurable concentration limits, so-called Derived Concentration Guidance Levels (DCGLs), should be developed for each radionuclide which is expected to be present in the site. Also, as they serve as a goal of decommissioning and direct dismantling and decontamination methods applicable to the site, DCGLs should be developed in the early phase of decommissioning. This paper describes how each overseas nuclear power plant developed its site-specific Soil DCGLs: what kind of post closure use of the site (scenario) was assumed and how the site-specific Soil DCGLs were calculated based on the scenario assumed for each plant. Through this, it is intended to derive lessons learned which will be instructive for future decommissioning of domestic nuclear power plants including Kori Unit 1. It is very important to have as good under-standing as possible of characteristics of the site by collection of relevant information and data in order to apply a scenario which is most foreseeable and plausible for a site to be decommissioned and to provide site-specific inputs to the calculation of the Soil DCGLs. These efforts will help to have not-overly conservative values for the Soil DCGLs, thus thereby reducing the costs and time needed for performing the decommissioning.

Environmental radiation monitoring around the nuclear facilities

International Nuclear Information System (INIS)

Lee, H.D.; Lee, Y.B.; Lee, W.Y.; Park, D.W.; Chung, B.G.

1980-01-01

For the KAERI site, various environmental samples were collected three times a month, and the natural environmental radiation levels were also measured at each sampling point. Measurements for gross alpha and beta radioactivities of the samples were routinely measured for all samples. Strontium-90 concentrations were also analysed for the fallout and air samples collected daily basis on the roof of the main building. Accumulated exposure including the possibility of determination of low level environmental radiation field by employing thermoluminescent dosimeter, CaSO 4 : Dsub(y)-0.4 teflon disc type, at 6 posts in on-site of the KAERI. As for Kori site, at 19 points of ON, OFF-site, and at the same time the environmental radiation exposure rate at each sampling point were measured. Several environmental samples such as surface soil, pine needles, water samples, milk sample and pasture samples were collected and analysed on a quarterly basis. As a result of the survey it can be said that no significant release of radiation to the environment due to the operations of nuclear facilities including research reactor at the KAERI and power reactor at the Kori has been found during the period of the survey and monitoring. (author)

Technical evaluation of the alternate to the keylock control to the bypass valves for the Davis-Besse nuclear power plant, Unit 1

International Nuclear Information System (INIS)

Ibarra, J.G.

1979-09-01

This report documents the technical evaluation of the alternate to the keylock control to the bypass valves for the Davis-Besse nuclear power plant, Unit 1. The review criteria are inferred from the NRC Reactor Safety Study (WASH-1400) and the Safety Evaluation Report for Davis-Besse. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

Report of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

International Nuclear Information System (INIS)

2001-12-01

In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The following findings and conclusion were made as the result of the present examination. (1) Wall thickness of the pipe was significantly reduced in the ruptured region. (2) Dimple pattern resulting from ductile fracture by shearing was observed in the fracture surfaces of nearly all of the pieces and no indication of fatigue crack growth was found. (3) Microstructure showed a typical carbon

Technical evaluation of RETS-required reports for Browns Ferry Nuclear Power Station, Units 1, 2, and 3, for 1983

International Nuclear Information System (INIS)

Young, T.E.; Magleby, E.H.

1985-01-01

A review was performed of reports required by federal regulations and the plant-specific radiological effluent technical specifications (RETS) for operations conducted at Tennessee Valley Authority's Browns Ferry Nuclear Station, Units 1, 2, and 3, during 1983. The two periodic reports reviewed were (a) the Effluents and Waste Disposal Semiannual Report, First Half 1983 and (b) the Effluents and Waste Disposal Semiannual Report, Second Half 1983. The principal review guidelines were the plant's specific RETs and NRC guidance given in NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants.'' The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines

Interim reliability-evaluation program: analysis of the Browns Ferry, Unit 1, nuclear plant. Appendix B - system descriptions and fault trees

International Nuclear Information System (INIS)

Mays, S.E.; Poloski, J.P.; Sullivan, W.H.; Trainer, J.E.; Bertucio, R.C.; Leahy, T.J.

1982-07-01

This report describes a risk study of the Browns Ferry, Unit 1, nuclear plant. The study is one of four such studies sponsored by the NRC Office of Research, Division of Risk Assessment, as part of its Interim Reliability Evaluation Program (IREP), Phase II. This report is contained in four volumes: a main report and three appendixes. Appendix B provides a description of Browns Ferry, Unit 1, plant systems and the failure evaluation of those systems as they apply to accidents at Browns Ferry. Information is presented concerning front-line system fault analysis; support system fault analysis; human error models and probabilities; and generic control circuit analyses

Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

International Nuclear Information System (INIS)

1984-02-01

Supplement 17 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and the previous supplements

A Study on Determination of Proper Pressurizer Level for Kori Unit 1

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang Sup; Song, Dong Soo [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1996-12-31

1. Determination of Proper Pressurizer Level. To determine operation level of pressurizer, LOFTRAN code is used with conservative model and assumption. 2. Performance Analysis. To simulate the plant, RETRAN computer code is used with realistic model and assumptions. The control and protection systems are fully credited. The turbine trip event is simulated at the condition of 47.62% of pressurizer level at full power. After that the same event as that with 55% of pressurizer level is simulated. And the FSAR requirements of pressurizer are verified with the new level setpoints. 3. Safety Analysis. As safety analyses, Loss of Normal Feedwater/Station Blackout which is significantly affected by the initial pressurizer water level is performed. Turbine Trip accident is also analyzed to verify if the peak primary side pressure is within the limit. LOFTRAN code is used with conservative mode and assumption. 3. Steam Generator Replacement. Relating to the steam generator replacement planed in 1998, safety analysis in terms of pressurizer level setpoint change. 4. Limit Condition for Operation. The LCO of pressurizer level is changed from 60% to 67.4% which is included pressurizer level uncertainty. (author). 13 refs., figs., tabs.

Safety Evaluation Report related to the full-term operating license for Millstone Nuclear Power Station, Unit No. 1 (Docket No. 50-245)

International Nuclear Information System (INIS)

1985-10-01

The Safety Evaluation Report for the full-term operating license application filed by the Connecticut Light and Power Company, the Hartford Electric Light Company, Western Massachusetts Electric Company and the Millstone Point Company [(now known as Connecticut Light and Power Company (CL and P) and Western Massachusetts Electric Company (WMECO) having authority to possess Millstone-1, 2, and 3, and the Northeast Nuclear Energy Company (NNECO) as the responsible entity for operation of the facilities)] for Millstone Nuclear Power Station Unit 1 has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in the town of Waterford, Connecticut. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can continue to be operated without endangering the health and safety of the public

Use of percentile rank sum method in identifying repetitive high occupational radiation dose jobs in a nuclear power plant

International Nuclear Information System (INIS)

Cho, Y.H.; Ko, H.S.; Kim, S.H.; Kang, C.S.; Moon, J.H.; Kim, K.D.

2004-01-01

The cost-effective reduction of occupational radiation dose (ORD) at a nuclear power plant could not be achieved without going through an extensive analysis of accumulated ORD data of existing plants. Through the data analysis, it is required to identify what are the jobs of repetitive high ORD at the nuclear power plant. In general the point value method commonly used, over-estimates the role of mean and median values to identify the high ORD jobs which can lead to misjudgment. In this study, Percentile Rank Sum Method (PRSM) is proposed to identify repetitive high ORD jobs, which is based on non-parametric statistical theory. As a case study, the method is applied to ORD data of maintenance and repair jobs at Kori units 3 and 4 that are pressurized water reactors with 950 MWe capacity and have been operated since 1986 and 1987, respectively in Korea. The results were verified and validated, and PRSM has been demonstrated to be an efficient method of analyzing the data. (authors)

78 FR 22347 - GPU Nuclear Inc., Three Mile Island Nuclear Power Station, Unit 2, Exemption From Certain...

Science.gov (United States)

2013-04-15

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-320; NRC-2013-0065] GPU Nuclear Inc., Three Mile Island Nuclear Power Station, Unit 2, Exemption From Certain Security Requirements AGENCY: Nuclear... and State Materials and Environmental Management Programs, U.S. Nuclear Regulatory Commission...

A pilot application of risk-based methods to establish in-service inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station

International Nuclear Information System (INIS)

Vo, T.; Gore, B.; Simonen, F.; Doctor, S.

1994-08-01

As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish in-service inspection plans for nuclear power plant components. This method uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot applications of this method. The specific systems addressed in this report are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants. To develop inspection plans, the acceptable level of risk from structural failure for important systems and components will be apportioned as a small fraction (i.e., 5%) of the total PRA-estimated risk for core damage. This process will determine target (acceptable) risk and target failure probability values for individual components. Inspection requirements will be set at levels to assure that acceptable failure probabilistics are maintained

78 FR 32278 - Vogtle Electric Generating Station, Units 3 and 4; Southern Nuclear Operating Company; Change to...

Science.gov (United States)

2013-05-29

... Generating Station, Units 3 and 4; Southern Nuclear Operating Company; Change to Information in Tier 1, Table... Nuclear Operating Company, Inc., and Georgia Power Company, Oglethorpe Power Corporation, Municipal... Table 3.3-1, ``Definition of Wall Thicknesses for Nuclear Island Buildings, Turbine Buildings, and Annex...

Experience with HEPA filters at United States nuclear installations

International Nuclear Information System (INIS)

Bellamy, R.R.

1977-01-01

Part 50 of Title 10 of the United States Code of Federal Regulations requires that a number of atmosphere cleanup systems be included in the design of commercial nuclear power plants to be licensed in the United States. These filtering systems are to contain high efficiency particulate air (HEPA) filters for removal of radioactive particulate matter generated during normal and accident conditions. Recommendations for the design, testing and maintenance of the filtering systems and HEPA filter components are contained in a number of United States Nuclear Regulatory Commission documents and industry standards. This paper will discuss this published guidance available to designers of filtering systems and the plant operators of U.S. commercial nuclear power plants. The paper will also present a survey of published reports of experience with HEPA filters, failures and possible causes for the failures, and other abnormal occurrences pertaining to HEPA filters installed in U.S. nuclear power installations. A discussion will be included of U.S. practices for qualification of HEPA filters before installation, and verification of continued performance capability at scheduled intervals during operation

Erie Nuclear Plant, Units 1 and 2. License application, preliminary safety analysis report, volume 1: chapter 1

International Nuclear Information System (INIS)

1977-01-01

Erie 1 and 2 reactors will be located on a site consisting of 1,740 acres in Berlin Township about 2.4 miles south of Lake Erie and 8.9 miles southeast of Sandusky, Ohio. Lake Erie will provide makeup water. The ultimate heat sink will be a 40 acre cooling reservoir shared by both units. Each unit has a proposed core thermal power level of 3,760 MW(t) and a nominal net capacity of about 1282 MW(e). Unit 1 is scheduled for commercial operation by 4/1/84 and Unit 2 by 4/1/86. Fuel will be Zircaloy-4 clad tubes containing cylindrical fuel pellets of UO 2 . B and W pressurized water reactor units to be employed are described in DOCKET-STN-50561

State nuclear initiatives in the United States

International Nuclear Information System (INIS)

Strauss, P.L.; Stoiber, C.R.

1977-01-01

The paper deals with State nuclear initiatives regarding the role of nuclear power in the energy future of the United States. The question of whether and under what circumstances nuclear facilities should be used to generate electricity was put to the popular vote in several States in 1976. Some general principles of Federal-State relations are discussed with specific reference to nuclear regulations. The initiative mechanism itself is described as well as its legal form and background. The parallel developments in the State and Federal legislative consideration of nuclear issues is reviewed and the suggested reasons for the defeat of the proposals in the seven States concerned are discussed. Finally, the author draws some conclusions on the effects of the 1976 initiatives on future decision-making in the US on energy policy in general and nuclear power in particular. (NEA) [fr

78 FR 4467 - UniStar Nuclear Energy, Combined License Application for Calvert Cliffs Power Plant, Unit 3...

Science.gov (United States)

2013-01-22

... NUCLEAR REGULATORY COMMISSION [Docket No. 52-016; NRC-2008-0250] UniStar Nuclear Energy, Combined License Application for Calvert Cliffs Power Plant, Unit 3, Exemption 1.0 Background UniStar Nuclear Energy (UNE), on behalf of Calvert Cliffs Nuclear Project, LLC and UniStar Nuclear Operating Services...

Structural review of the Palisades Nuclear Power Plant Unit 1 containment structure under combined loads for the Systematic Evaluation Program

International Nuclear Information System (INIS)

Liaw, C.Y.; Debeling, A.; Tsai, N.C.

1981-12-01

A structural reassessment of the containment structure of the Palisades Nuclear Power Plant Unit 1 was performed for the Nuclear Regulatory Commission as part of the Systematic Evaluation Program. Conclusions about the ability of the containment structure to withstand the Abnormal/Extreme Environment are presented. The reassessment focused mainly on the overall structural integrity of the containment building for the Abnormal/Extreme Environment. In this case, the Abnormal Environmental condition is caused by the worst case of either a Loss-of-Coolant Accident or a main steam line break. The Extreme Environmental condition is the Safe Shutdown Earthquake

Development and application of an entrainment model for the PWR U-tube steam generators for main steam line break analysis

International Nuclear Information System (INIS)

Song, Dong-Soo; Park, Young-Chan

2004-01-01

The purpose of this paper is to present the analyses that were performed to develop and use an entrainment model for pressurized water reactor U-tube steam generators (SG) for main steam line break (MSLB) analyses. The entrainment model was developed using the RETRAN-3D computer program, and the model was benchmarked against experimental data of moisture carryover during a simulated MSLB accident. The application methodology was also developed to incorporate into the MSLB mass and energy release calculations for Kori Unit 1. This methodology utilizes LOFTRAN and RETRAN-3D codes in an iterative sequence of cases in which the LOFTRAN nuclear steam supply system model provides boundary conditions for the RETRAN-3D broken loop steam generator model, and the RETRAN-3D model provides the entrainment data that is input back into the LOFTRAN model. FORTRAN programs were developed to facilitate the sequencing of these iterative calculations. As a result of applying the entrainment model to Kori Unit 1, the temperature calculated inside Containment during MSLB accident using the CONTEMP-LT computer program decreased by about 25degC. Consequently this entrainment model provides a significant benefit by decreasing the temperature envelop for environment qualification as well as decreasing the peak Containment pressure. (author)

Safety Evaluation Report on Tennessee Valley Authority: Browns Ferry Nuclear Performance Plan: Browns Ferry Unit 2 restart

International Nuclear Information System (INIS)

1989-04-01

This safety evaluation report (SER) on the information submitted by the Tennessee Valley Authority (TVA) in its Nuclear Performance Plan, through Revision 2, for the Browns Ferry Nuclear Power Station and in supporting documents has been prepared by the US Nuclear Regulatory Commission staff. The plan addresses the plant-specific concerns requiring resolution before startup of Unit 2. The staff will inspect implementation of those programs. Where systems are common to Units 1 and 2 or to Units 2 and 3, the staff safety evaluations of those systems are included herein. 3 refs

Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

International Nuclear Information System (INIS)

1984-06-01

Supplement 23 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses the applicant's requests for approval of 22 deviations from the requirements of Section III.G of Appendix R of Title 10 of the Code of Federal Regulations Part 50

Development of a seismic damage assessment program for nuclear power plant structures

Energy Technology Data Exchange (ETDEWEB)

Koh, Hyun Moo; Cho, Ho Hyun; Cho, Yang Hui [Seoul National Univ., Seoul (Korea, Republic of)] (and others)

2000-12-15

Some of nuclear power plants operating currently in Korea have been passed about 20 years after construction. Moreover, in the case of KORI I the service year is over 20 years, so their abilities are different from initial abilities. Also, earthquake outbreak increase, our country is not safe area for earthquake. Therefore, need is to guarantee the safety of these power plant structures against seismic accident, to decide to maintain them operational and to obtain data relative to maintenance/repair. Such objectives can be reached by damage assessment using inelastic seismic analysis considering aging degradation. It appears to be more important particularly for the structure enclosing the nuclear reactor that must absolutely protect against any radioactive leakage. Actually, the tendency of the technical world, led by the OECD/NEA, BNL in the United States, CEA in France and IAEA, is to develop researches or programs to assess the seismic safety considering aging degradation of operating nuclear power plants. Regard to the above-mentioned international technical trend, a technology to establish inelastic seismic analysis considering aging degradation so as to assess damage level and seismic safety margin appears to be necessary. Damage assessment and prediction system to grasp in real-time the actual seismic resistance capacity and damage level by 3-dimensional graphic representations are also required.

Pilot Study for Maintenance Rule at KSNP

International Nuclear Information System (INIS)

Choi, Kwang Hee; Jeong, Hyeon Jong; Jee, Moon Hak; Hong, Sung Yull

2005-01-01

Maintenance Rule (MR), which was developed to monitor the effectiveness of maintenance in a nuclear power plant (NPP), has been received as highly successful program by and large since its implementation in 1996 in the United States. Korea has initiated two pilot programs to implement the Maintenance Rule program in 2003. Selected plants for the pilot implementation are Kori 3 and 4 units and Ulchin 3 and 4 units, where Kori 3 and 4 units are Westinghouse units and Ulchin 3 and 4 units are Korean Standardized Nuclear Power (KSNP) Plant units. This paper describes the results of each key tasks completed to date and insights gained from pilot study on the KSNP units. Currently, Scoping of the functions of maintenance rule and determination of safety significance level have been completed during first year. As first task, total 607 functions were identified and defined by detailed function analysis on 135 systems that cover all plant systems. About 55% of total functions are selected as within the scope of maintenance rule. Among these inscoped functions, 56% of scoped functions are safety related and 44% are non-safety related functions. Evaluation of safety significance for each function was determined by expert panel consist of eight experts in field of plant maintenance, operation, PSA, work schedule and system engineers. As a result, about 46% of functions were determined to be high safety significant functions and rest of the functions were classified as low safety significant. The remaining tasks that are included determination of performance criteria and preparation of implementing guideline will be performed in following years

Nuclear safety risk control in the outage of CANDU unit

International Nuclear Information System (INIS)

Wu Mingliang; Zheng Jianhua

2014-01-01

Nuclear fuel remains in the core during the outage of CANDU unit, but there are still nuclear safety risks such as reactor accidental criticality, fuel element failure due to inability to properly remove residual heat. Furthermore, these risks are aggravated by the weakening plant system configuration and multiple cross operations during the outage. This paper analyzes the phases where there are potential nuclear safety risks on the basis of the typical critical path arrangement of the outage of Qinshan NPP 3 and introduces a series of CANDU-specific risk control measures taken during the past plant outages to ensure nuclear safety during the unit outage. (authors)

Subsidence analysis Forsmark nuclear power plant - unit 1

International Nuclear Information System (INIS)

Bono, Nancy; Fredriksson, Anders; Maersk Hansen, Lars

2010-12-01

On behalf of SKB, Golder Associates Ltd carried out a risk analysis of subsidence during Forsmark nuclear power plant in the construction of the final repository for spent nuclear fuel near and below existing reactors. Specifically, the effect of horizontal cracks have been studied

Structural design of the turbine building of Angra Nuclear Power Station, Unit 1

International Nuclear Information System (INIS)

Varella, L.N.; Reis, F.J.C.; Jurkiewicz, W.J.

1978-01-01

The Turbine Building of the Angra Nuclear Power Plant, Unit 1, and particularly its structure and structural design are described. The Turbine Building, as far as its structure is concerned, deviates from the standard structure of any turbine building due to the fact that huge ducts are provided in the foundation mat as to accomodate the circulating water system. This aspect and the fact that the building is founded upon a very deep strata of compacted and controlled fill, makes out of the building structure 'a concrete ship floating in the sea of sand', and by the same reason presents by itself an interesting structure, worth to be known to all engineers involved in design of power plants. This pape, suplemented by a few slides shown during presentation of the paper at the conference, covers the subject mainly from the designers' point of view. (Author)

Cernavoda NPP Unit 2 and Romanian nuclear industry

International Nuclear Information System (INIS)

Mocanu Horia

2001-01-01

On 18 May 2001, in the presence of Mr. Adrian Nastase, the Prime-Minister of Romania, the presidents of AECL, ANSALDO and Director General of SN Nuclearelectrica, the commercial and management contract for completing the Cernavoda NPP Unit 2 was signed. This document stipulates the goal and the partners' commitments, leadership organization, the SN Nuclearelectrica's control of the Budget, costs for the technical assistance (around 180 specialists from abroad), as well as the costs of equipment supplied from Canada and Italy. Services and equipment supplied by Canada and Italy amounts up to around USD 300 millions. Efforts are currently undertaken to obtain a loan of USD 300 millions from EURATOM, beginning from 2003. An auction process, implying around 10 companies, is underway and by the completion of the process, in February 2002, the practical delivery of equipment will start. The so-far invested capital amounts around USD 650 millions while the capital funds remaining to be invested amounts up to about USD 689 millions. From the latter figure, around USD 100 millions represent the costs for heavy water and the initial nuclear fuel charging. The personnel dynamics is presented as well as problems relating with recruitment and salary policy. Romanian nuclear industry is engaged for supply of a series of important components. General Turbo SA, supplied already components of some tens USD millions for the turbogenerator complex. PETROTUB company from Roman, Romania supplied already one thousand tones of non-nuclear carbon steel tubing valued at about USD 300 millions. ARIO, Bistrita, Romania, has signed contracts valued at about USD 400,000 for non-nuclear reinforcing materials. Other companies like AVERSA SA and Ventilatorul SA supplied reliable equipment for Unit 1 and will continue to do the same for Unit 2. Contracts of over one million USD are carried on with VULCAN for carbon steel fittings and with TITAN Nuclear Equipment for components of the fueling

Changing Perceptions of Nuclear Power in The United States

International Nuclear Information System (INIS)

Taylor, John

1989-01-01

Although many new nuclear power plants have been brought on line in that time, resulting in a capacity of 110 plants with operating permits and another twelve in the last stages of completion, all of these plants were authorized before 1978. The fundamental reason for this moratorium in new orders was the precipitous reduction in electricity demand, arising from the OPEC embargo and Iran revolution, which created excess electric capacity throughout the United States. In fact, many nuclear and coal plants were cancelled to minimize the over capacity problem and no large base load generating units have been ordered of any kind in the past decade. So the 'moratorium' is not really unique to nuclear power. Progress, coupled with increased awareness that nuclear power is one of the keys to solving atmospheric environmental problems, will swing political and public acceptance back to being favorable. Successful progress in these matters will be of benefit to public acceptance around the world and, conversely, serious technical difficulties, particularly entailing any major incident with a nuclear power plants anywhere in the world, will adversely affect the improvement in political and public acceptance in the United States. It is vitally important, therefore, that we continue to further enhance international cooperation in nuclear power. We are pleased the Korea Electric power Corporation and the Korea Advanced Energy Research Institute are participating in EPRI development programs, and hope that cooperation will increase in the future. We're most encouraged by the formation of the World Association of Nuclear Operators, which will be initiated in Moscow next month. The nuclear electric utilities and their governments around the world, the International Atomic Energy Agency, and the Nuclear Energy Agency of OECD should be commended for their initiative in international cooperation

The United States nuclear plant reliability data program: Its description and status

International Nuclear Information System (INIS)

Wise, M.J.

1975-01-01

The American National Standards Institute Subcommittee N18-20 has developed and implemented the United States Nuclear Plant Reliability Data System (NPRDS). The NPRDS is designed to accumulate, store, analyse, and report reliability and failure statistics on systems and components of nuclear power plants related to nuclear safety. Input data to the NPRDS consist of engineering, operating, and failure information submitted on a voluntary basis by participating utilities. Prior to entry into the computerized data base, the data are thoroughly checked for accuracy by both the submitting organizations and the NPRDS operating contractor. The data base is the source of various periodic output reports to the nuclear power industry and is utilized to produce special reports upon request. The present data base represents data accumulated from about thirty nuclear units with additional units expected to begin submitting data immediately. The objective is to have essentially all operating nuclear units in the United States of America participating in the program by the end of 1975. The first NPRDS annual reports containing meaningful reliability and failure statistics are expected to be produced following the end of 1975. (author)

Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

International Nuclear Information System (INIS)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

1994-01-01

The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

Improvements in Logic Diagram of Computerized Procedure System of APR1400

Energy Technology Data Exchange (ETDEWEB)

Kang, Sungkweon; Seong, Nokyu [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Computerized Procedure System (CPS) has been improved since it is installed in Shin-Kori 3 and 4 Nuclear Power Plants. It is one of operating support systems of digital Main Control Room (MCR) and provides many functions to operators in executing the procedure. CPS can effectively remove the human errors by supporting the procedure flow and logic diagram. This paper describes the logic diagram of CPS of reference power plant and shows the improved logic diagram of CPS of Shin-Kori unit 5 and 6. This paper describes the current logic diagram of CPS and suggests improved design for logic diagram. The improved logic diagram shall be validated through human factors engineering verification and validation. The improved design will help operators execute the computerized procedure fast and remove the human error.

Principles and foundation: national standards on quantities and units in nuclear science field

International Nuclear Information System (INIS)

Chen Lishu

1993-11-01

The main contents of National Standards on Quantities and units of atomic and nuclear physics (GB 3102.9) and Quantities and Units of nuclear reactions and ionizing radiations (GB 310.10) are presented in which most important quantities with their symbols and definitions in the nuclear scientific field are given. The principles and foundation, including the International System of Units (SI) and its application to the nuclear scientific field, in the setting of the National Standards are explained

New steam generators slated for nuclear units

International Nuclear Information System (INIS)

Anon.

1992-01-01

This article is a brief discussion of Duke Power's plans to replace steam generators at its McGuire and Catawba nuclear units. A letter of intent to purchase (from Babcock and Wilcox) the 12 Westinghouse steam generators has been signed, but no constructor has been selected at this time. This action is brought about by the failures of more than 3000 tubes in these units

Safety evaluation report related to the operation of Sequoyah Nuclear Plant, Units 1 and 2, Docket Nos. 50-327 and 50-328, Tennessee Valley Authority

International Nuclear Information System (INIS)

1982-12-01

Supplement No. 6 to the Safety Evaluation Report (SER) related to the operation of the Tennessee Valley Authority's Sequoyah Nuclear Plant, Units 1 and 2, located in Hamilton County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The purpose of this supplement is to update the staff's evaluations of the issues related to the hydrogen mitigation system identified in the SER and previous supplements as needing resolution

Reorganization of Kepco and lifetime management status in Korea

International Nuclear Information System (INIS)

Chung, K.N.

2000-01-01

To date the Korea Electric Power Corporation ('KEPCO') has been a vertically integrated public utility. It has been the only company engaged in the generation, transmission and distribution of electricity with ownership of 94% of the total electricity generating capacity in Korea. The restructuring plan published last year by the Government involved a gradual transition to wholesale competition over four years, with the introduction of retail competition for small consumers after 2009. In the initial period, KEPCO's generation assets will be divided into six generation companies for divestment and/or privatisation. Subsequently, distribution assets will also be divided into a number of companies. KEPCO will remain as a national transmission company offering access to the Grid on non-discriminatory terms to promote the electricity market. Although many details will require further clarification in coming months, the Korean electricity supply industry will be moving towards a competitive market structure. The impending restructuring will force the nuclear sector to reassess the way they do business. The nuclear sector consisting of 16 units in operation and 4 units under construction has competitiveness over the conventional electricity resource. To survive in the competitive market and the deregulated industry, it is important to reduce the capital expenditure as well as fuel and O and M cost. It was already known in other countries that it is more efficient for competition in the market to better utilize the operating plants rather than to reduce the construction cost of new plants. In this context, consideration of lifetime extension for older nuclear units may naturally occur in parallel with elaborate evaluations to make massive capital investments into new power plants. Therefore, the first PLIM programme is applied for Kori 1, the oldest nuclear unit in Korea, which is scheduled to retire in 2008. The first phase of the programme identified that the life

Unconventional wisdom [Entergy's economic management of five disparate nuclear units

International Nuclear Information System (INIS)

Wood, J.

1996-01-01

The conventional wisdom is that a competitive nuclear power programme requires standard plants, multi-unit sites and economies of scale. Entergy Operations, Inc (EOI) in the USA has five nuclear units of two reactor types, acquired from five vendors and located on four sites separated by hundreds of miles. EOI set two bench marks in its aim to deliver electricity at the market price from each of its nuclear units while meeting the costs of O and M, fuel, capital investment and overheads. The first was to become the best operator of each type of its differently designed units. The second was to combine the sense of ownership and expertise of single unit operators with the flexibility and efficiency of multi-unit sites. In order to achieve this, EOI aims to operate the five units as though they were on one site. Central to this plan is a set of peer groups, each of which deals with an aspect of plant operation, sharing good practice across all the units. Building on this, increasingly during outages, employees are brought in from other units rather than employing outside contractors. Although EOI employees are more expensive on an hourly basis they are two to three times more efficient than outside contractors. The peer group concept is also leading to rationalisation of expertise as groups become identified with particular specialisations which back-up teams working on other units. (U.K.)

Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 33

International Nuclear Information System (INIS)

1986-05-01

Supplement 33 to the Safety Evaluation Report for the Pacific Gas and Electric Company's Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. The supplement reports on the status of the staff's investigation, inspection, and evaluation of allegations and concerns that have been identified to the NRC through March 1986. The report includes a complete listing of all allegations and concerns, indicating the status of their resolution. The NRC staff concludes that the technical issues raised in the allegations with regard to the design, construction, and safe operation of Diablo Canyon Units 1 and 2 have been satisfactorily resolved and no further action is required

Sweden, United States and nuclear energy. The establishment of a Swedish nuclear materials control 1945-1995

International Nuclear Information System (INIS)

Jonter, T.

1999-05-01

This report deals mainly with the United States nuclear energy policy towards Sweden 1945-1960. Although Sweden contained rich uranium deposits and retained high competence in the natural sciences and technology, the country had to cooperate with other nations in order to develop the nuclear energy. Besides developing the civil use of nuclear power, the Swedish political elite also had plans to start a nuclear weapons programme. From the beginning of the 1950s up to 1968, when the Swedish parliament decided to sign the non-proliferation treaty, the issue was widely debated. In this report, American policy is analyzed in two periods. In the first period, 1945-1953, the most important aim was to prevent Sweden from acquiring nuclear materials, technical know-how, and advanced equipment which could be used in the production of nuclear weapons. The Swedish research projects were designed to contain both a civil and military use of nuclear energy. The first priority of the American administration was to discourage the Swedes from exploiting their uranium deposits, especially for military purposes. In the next period, 1953-1960, the American policy was characterized by extended aid to the development of the Swedish energy programme. Through the 'Atoms for Peace'-programme, the Swedish actors now received previously classified technical information and nuclear materials. Swedish companies and research centers could now buy enriched uranium and advanced equipment from the United States. This nuclear trade was, however, controlled by the American Atomic Energy Commission (AEC). The American help was shaped to prevent the Swedes from developing nuclear weapons capability. From mid-50s Swedish politicians and defence experts realised that a national production of nuclear bombs would cost much more money than was supposed 4-5 years earlier. As a consequence, Swedish officials started to explore the possibilities of acquiring nuclear weapons from United States. The American

Calculation Method of Steam Generator Level for swelling and shrinking effects in YGN 1/2 Simulator

International Nuclear Information System (INIS)

Hwang, Do Hyun; Seo, In Yong; Park, Weon Seo; Suh, Jae Seung

2007-01-01

In August 2006, the development of new simulator for YGN 1/2 Simulator was completed. The NSSS (Nuclear Steam Supply System) T/H(Thermal- Hydraulic) module in this simulator was developed with ARTS code based on RETRAN, which is a best estimate thermal-hydraulic code designed to analyze several operational transients by EPRI(Electric Power Research Institute). RETRAN, however, has some limitations in real-time calculation capability and its robustness to be used in the simulator for some transient conditions. To overcome these limitations, its robustness and real-time calculation capability have been improved with simplifications and removing of discontinuities of the physical correlations of the RETRAN code. And some supplements are also developed to extend its simulation scope of the ARTS code. In comparison to KNPEC(Kori Nuclear Power Education Center) no.2 simulator, the simulator based on Younggwang Unit 1 developed in the year 2001, the ARTS code was upgraded that it extended its calculating region to the steam line and common header before turbine while it had calculated to the steam generator exit before steam line in KNPEC no.2 simulator. Consequently, the number of volume and fill/normal junction in ARTS nodalization increased to 109 and 174 from 62 and 125, respectively

The Evolution of India's Nuclear Program: Implications for the United States

National Research Council Canada - National Science Library

Creasman, David J

2008-01-01

Since India began developing its nuclear program it has continually encountered issues with the United States and other nuclearized countries over whether India should be able to establish a nuclear...

Choices: A Unit on Conflict and Nuclear War.

Science.gov (United States)

Massachusetts Teachers Association, Boston.

Ten lessons on the evolution of the nuclear arms race, the nature and consequences of using nuclear weapons, and new ways that conflicts among nations might be resolved are presented for the junior high school level. The unit contains age-appropriate materials to equip students with skills and knowledge to understand what choices can be made to…