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Sample records for bwr pilot plant

  1. Development of next BWR plant

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

  3. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  4. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  5. BWR control rod drive scram pilot valve monitoring system

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1984-01-01

    The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechancial works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the ''insert'' side of the control rod piston and vents the ''withdraw'' side of the piston causing the rods to insert during a scam. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a ''half scram'', a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

  6. BWR control rod drive scram pilot valve monitoring program

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1986-01-01

    The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechanical works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the insert side of the control rod piston and vents the withdraw side of the piston causing the rods to insert during a scram. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a half scram, a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

  7. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  8. Managing BWR plant life extension

    International Nuclear Information System (INIS)

    Ianni, P.W.; Kiss, E.

    1985-01-01

    Recent studies have confirmed that extending the useful life of a large nuclear plant can be justified with very high cost benefit ratio. In turn, experience with large power plant systems and equipment has shown that a well-integrated and -managed plan is essential in order to achieve potential economic benefits. Consequently, General Electric's efforts have been directed at establishing a life extension plan that considers alternative options and cost-effective steps that can be taken in early life, those appropriate during middle life, and those required in late life. This paper briefly describes an approach designed to provide the plant owner a maximum of flexibility in developing a life extension plan

  9. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  10. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  11. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  12. Method of operating BWR type power plants

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

  13. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  14. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  15. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  16. Construction techniques and management methods for BWR plants

    International Nuclear Information System (INIS)

    Shimizu, Yohji; Tateishi, Mizuo; Hayashi, Yoshishige

    1989-01-01

    Toshiba is constantly striving for safer and more efficient plant construction to realize high-quality BWR plants within a short construction period. To achieve these aims, Toshiba has developed and improved a large number of construction techniques and construction management methods. In the area of installation, various techniques have been applied such as the modularization of piping and equipment, shop installation of reactor internals, etc. Further, installation management has been upgraded by the use of pre-installation review programs, the development of installation control systems, etc. For commissioning, improvements in commissioning management have been achieved through the use of computer systems, and testing methods have also been upgraded by the development of computer systems for the recording and analysis of test data and the automatic adjustment of controllers in the main control system of the BWR. This paper outlines these construction techniques and management methods. (author)

  17. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

  18. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  19. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  20. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  1. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    characteristics. In this paper, related publications will be reviewed to enhance the understanding of parametric effects to the system stability. Subsequently, core stability analysis for the Kuosheng Nuclear Power Plant (BWR6) will be conducted, and then parallel verification against vendor's results will be performed also. We will summarize our efforts in using LAPUR5 methodologies for BWR6 core stability analysis and share our experiences among international communities. (authors)

  2. Availability analysis of United States BWR IV electrical generation plants

    International Nuclear Information System (INIS)

    Renick, D.H.; Li, F.; Todreas, N.E.

    1998-01-01

    Availability, as quantified by power output levels, from all active U.S. BWR IV plants were analyzed over a seven and a half year period to determine the operational characteristics of these plants throughout an operating cycle. The operational data were examined for infant mortality, end of cycle decreased availability, and seasonal availability variations. Scheduled outages were also examined to determine the industry's current approach to planning maintenance outages. The results of this study show that nuclear power plants do suffer significant infant mortality following a refueling outage. And while they do not suffer an end of cycle decrease in availability, a mid-cycle period of decreased availability is evident. This period of decreased availability is due to a combination of increased forced unavailability and seasonally scheduled maintenance and refueling outages. These findings form the start of a rational approach to increasing plant availability. (author)

  3. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  4. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  5. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Monta, K.; Itoh, J.

    1992-01-01

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  6. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    1999-05-01

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  7. Dose rate reduction method for NMCA applied BWR plants

    International Nuclear Information System (INIS)

    Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

    2012-09-01

    BRAC (BWR Radiation Assessment and Control) dose rate is used as an indicator of the incorporation of activated corrosion by products into BWR recirculation piping, which is known to be a significant contributor to dose rate received by workers during refueling outages. In order to reduce radiation exposure of the workers during the outage, it is desirable to keep BRAC dose rates as low as possible. After HWC was adopted to reduce IGSCC, a BRAC dose rate increase was observed in many plants. As a countermeasure to these rapid dose rate increases under HWC conditions, Zn injection was widely adopted in United States and Europe resulting in a reduction of BRAC dose rates. However, BRAC dose rates in several plants remain high, prompting the industry to continue to investigate methods to achieve further reductions. In recent years a large portion of the BWR fleet has adopted NMCA (NobleChem TM ) to enhance the hydrogen injection effect to suppress SCC. After NMCA, especially OLNC (On-Line NobleChem TM ), BRAC dose rates were observed to decrease. In some OLNC applied BWR plants this reduction was observed year after year to reach a new reduced equilibrium level. This dose rate reduction trends suggest the potential dose reduction might be obtained by the combination of Pt and Zn injection. So, laboratory experiments and in-plant tests were carried out to evaluate the effect of Pt and Zn on Co-60 deposition behaviour. Firstly, laboratory experiments were conducted to study the effect of noble metal deposition on Co deposition on stainless steel surfaces. Polished type 316 stainless steel coupons were prepared and some of them were OLNC treated in the test loop before the Co deposition test. Water chemistry conditions to simulate HWC were as follows: Dissolved oxygen, hydrogen and hydrogen peroxide were below 5 ppb, 100 ppb and 0 ppb (no addition), respectively. Zn was injected to target a concentration of 5 ppb. The test was conducted up to 1500 hours at 553 K. Test

  8. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  9. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  10. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  11. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  12. Two-loop feed water control system in BWR plants

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1982-01-01

    In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

  13. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  14. Description of the power plant model BWR-plasim outlined for the Barsebaeck 2 plant

    International Nuclear Information System (INIS)

    Christensen, P. la Cour.

    1979-08-01

    A description is given of a BWR power plant model outlined for the Barsebaeck 2 plant with data placed at our disposal by the Swedish Power Company Sydkraft A/B. The basic operations are derived and simplifications discussed. The model is implemented with a simulation system DYSYS which assures reliable solutions and easy programming. Emphasis has been placed on the models versatility and flexibility so new features are easy to incorporate. The model may be used for transient calculations for both normal plant conditions and for abnormal occurences as well as for control system studies. (author)

  15. Kinetic analyses on startup and shutdown chemistry of BWR plant

    International Nuclear Information System (INIS)

    Domae, Masafumi; Fujiwara, Kazutoshi; Inagaki, Hiromitsu

    2012-09-01

    During startup and shutdown of Boiling Water Reactor (BWR) plants, temperature and dissolved oxygen (DO) concentration of reactor water change in a wide range. The changes result in variation of conductivity and pH of the reactor water. It has been speculated that the water chemistry change is due to dissolution of the oxides on fuel claddings and structural materials. However, detailed mechanism is not known. In the present paper, trend of recent water chemistry in several BWR plants during startup and shutdown is presented. Conductivity and pH are convenient indication of coolant purity. We tried to clarify the mechanism of the change in the conductivity and the pH value during startup and shutdown, based on the water chemistry data measured. In the water chemistry data, change in chromate concentration and Ni 2+ concentration is rather large. It is assumed that change in the chromate concentration and the Ni 2+ concentration results in the time variation of the conductivity and the pH value. It is reasonable to consider that the increase in the chromate concentration and the Ni 2+ concentration is ascribed to dissolution of Cr oxides and Ni oxides, respectively. A model of dissolution of the Cr oxides and the Ni oxides is proposed. A concept of finite inventory of the Cr oxides and the Ni oxides in the coolant system is introduced. The model is as follows. Chromate is generated by oxidation of the Cr oxides and the Cr dissolution rate depends on the DO concentration. The dissolution rate of chromate is in proportion to DO concentration, the inventory of Cr and difference between solubility limit and the chromate concentration. On the other hand, Ni 2+ is formed by dissolution of the Ni oxides, and DO is not necessary in this process. The dissolution rate of Ni 2+ is in proportion to the inventory of Ni and difference between solubility limit and the Ni 2+ concentration. Coolant is continuously purified, and the chromate concentration and the Ni 2+ concentration

  16. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    Andrade, G.G. de.

    1973-01-01

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  17. Power plant design: ESBWR - the latest passive BWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

    1997-01-01

    When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

  18. Automatic refueling platform and CRD remote handling device for BWR plant

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Takagi, Kaoru

    1978-01-01

    In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

  19. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  20. Process Experimental Pilot Plant

    International Nuclear Information System (INIS)

    Henze, H.

    1986-01-01

    The Process Experimental Pilot Plant (PREPP) at the Idaho National Engineering Laboratory (INEL) was built to convert transuranic contaminated solid waste into a form acceptable for disposal at the Waste Isolation Pilot Plant (WIPP), located near Carlsbad, New Mexico. There are about 2.0 million cubic ft of transuranic waste stored at the Transuranic Storage Area of the INEL's Radioactive Waste Management Complex (RWMC). The Stored Waste Examination Pilot Plant (SWEPP) located at the RWMC will examine this stored transuranic waste to determine if the waste is acceptable for direct shipment to and storage at WIPP, or if it requires shipment to PREPP for processing before shipment to WIPP. The PREPP process shreds the waste, incinerates the shredded waste, and cements (grouts) the shredded incinerated waste in new 55-gal drums. Unshreddable items are repackaged and returned to SWEPP. The process off-gas is cleaned prior to its discharge to the atmosphere, and complies with the effluent standards of the State of Idaho, EPA, and DOE. Waste liquid generated is used in the grouting operation

  1. Application of process computers and colour CRT displays in the plant control room of a BWR

    International Nuclear Information System (INIS)

    Itoh, M.; Hayakawa, H.; Kawahara, H.; Neda, T.; Wakabayashi, Y.

    1983-01-01

    The recent application of a CRT display system in an 1100-MW(e) BWR plant control room and the design features of a new control room whose installation is planned for the next generation are discussed. As reactor unit capacity and the need for plant safety and reliability continue to increase, instrumentation and control equipment is growing in number and complexity. In consequence, control and supervision of plant operations require improvement. Thus, because of recent progress in the field of process computers and display equipment (colour CRTs), efficient improvements of the control room are under way in the Japanese BWR plant. In the recently constructed BWR plant (1100 MW(e)), five CRTs on the bench board and two process computers were additionally installed in the control room during the construction stage to improve plant control and supervisory functions by implementing the lessons learned from the Three Mile Island incident. The major functions of the new computers and display systems are to show integrated graphic displays of the plant status, to monitor the standby condition of the safety system, to show the condition of the integrated alarm system, etc. In practice, in the actual plant, this newly installed system performs well. On the basis of the experience gained in these activities, a new computerized control and monitoring system is now being designed for subsequent domestic BWR plants. This advanced system will incorporate not only the functions already mentioned, but also a surveillance guide system and plant automation. For future plants, a diagnostic system and an instructional system that can analyse a disturbance and give operational guidance to the plant operator are being developed in a government-sponsored programme. (author)

  2. Development and recent trend of design of BWR nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kani, J [Tokyo Shibaura Electric Co. Ltd., Kawasaki, Kanagawa (Japan)

    1977-11-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation.

  3. Development and recent trend of disign of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Kani, Jiro

    1977-01-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

  4. Pilot plant study

    International Nuclear Information System (INIS)

    Morris, M.E.

    1978-01-01

    Sandia Laboratories undertook the design and fabrication of an 8 ton/day dry sewage sludge irradiatior. The facility is intended (1) to function as a high-gamma-dose rate research facility; (2) to be a testbed for the unique electrical and mechanical components to be used in larger facilities; (3) to fulfill the formal requirements of a pilot plant so that design and construction of a demonstration facility could proceed; and (4) to provide accurate data base on construction and operating experience for the Environmental Impact Assessment (EIA), the Safety Analysis Report (SAR), and the cost analyses for a larger facility. The facility and its component systems are described in detail

  5. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

  6. Development of a computerized operator support system for BWR power plant

    International Nuclear Information System (INIS)

    Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

    1985-01-01

    A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

  7. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  8. Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke [Energy Research Lab., Ibaraki (Japan); Ohsumi, Katsumi; Takashima, Yoshie [Hitachi Works, Ibaraki (Japan)

    1995-03-01

    Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increase must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.

  9. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

    2004-01-01

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  10. Dry well cooling systems in BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Hanamura, Ikuo; Tada, Kenji.

    1986-01-01

    Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

  11. Turbine protecting device in a BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kasuga, Hajime; Oka, Yoko.

    1984-01-01

    Purpose: To prevent highly humid steams from flowing into the turbine upon abnormal reduction in the reactor water level in order to ensure the turbine soundness, as well as in order to trip the turbine with no undesired effect on the reactor. Constitution: A protection device comprising a judging device and a timer are disposed in a BWR type reactor, in order to control a water level signal from a reactor water level gage. If the reactor water level is reduced during rated power operation, steams are kept to be generated due to decay heat although reactor is scramed. When a signal from the reactor water level detector is inputted to the protection device, a trip signal is outputted by way of a judging device after 15 second by means of the timer, when the main steam check valve is closed to trip the turbine. With this delay of time, abrupt increase in the pressure of the reactor due to sudden shutdown can be prevented. (Nakamoto, H)

  12. Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

    1986-01-01

    The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

  13. The Marcoule pilot plant

    International Nuclear Information System (INIS)

    Faugeras, P.; Calame Longjean, A.; Le Bouhellec, J.; Revol, G.

    1986-06-01

    The Marcoule spent fuel reprocessing pilot facility was built in 1960-1961 for extended testing of the PUREX process with various types of fuel under conditions similar to those encountered in a production plant. Extensive modification work was undertaken on the facility in 1983 in the scope of the TOR project, designed with the following objectives: - increase the throughput capacity to at least 5 metric tons of PHENIX equivalent fuel per year, - extend equipment and process R and D capability, - improve job safety by maximum use of remote handling facilities, - maximize waste conditioning treatments to produce waste forms suitable for direct storage, - provide a true industrial process demonstration in continuous operation under centralized control using computerized procedures. The redesigned plant is scheduled to begin operation during the second half of 1986. The proximity of the Industrial Prototypes Service and the ATALANTE radiochemical research laboratory scheduled to begin operation in 1990, will provide a synergistic environment in which R and D program may be carried out under exceptional conditions

  14. Operation status display and monitoring system for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Wakabayashi, Yasuo; Hayakawa, Hiroyasu; Kawamura, Atsuo; Kaneda, Mitsunori.

    1982-01-01

    Lately, the development of the system has been made for BWR plants, which monitors the operating status not only in normal operation but also in abnormal state and also for plant safety. Recently, the improvement of man-machine interface has been tried through the practical use of technique which displays data collectively on a CRT screen relating them mutually. As one of those results, the practical use of an electronic computer and color CRT display for No. 1 unit in the Fukushima No. 2 Nuclear Power Station (2F-1), Tokyo Electric Power Co., is described. Also, new centralized control panels containing such systems were used for the 1100 MWe BWR nuclear power plants now under construction, No. 3 unit of the Fukushima No. 2 Power Station and No. 1 unit of Kashiwazaki-Kariwa Nuclear Power Station (2F-3 and K-1, respectively). The display and monitoring system in 2F-1 plant is the first one in which a computer and color CRTs were practically employed for a BWR plant in Japan, and already in commercial operation. The advanced operating status monitoring system, to which the result of evaluation of the above system was added, was incorporated in the new centralized control panels presently under production for 2F-3 and K-1 plants. The outline of the system, the functions of an electronic computer, plant operating status monitor, surveillance test guide, the automation of plant operation and auxiliary operation guide are reported for these advanced monitoring system. It was confirmed that these systems are useful means to improve the man-machine communication for plant operation minitoring. (Wakatsuki, Y.)

  15. Kvanefjeld refinery pilot plant operations

    International Nuclear Information System (INIS)

    Krebs, Damien; Furfaro, Domenic

    2016-01-01

    Greenland Minerals and Energy is a junior project development company which is listed on the Australian Stock Exchange (asx:GGG). It is developing the Kvanefjeld rare earth and uranium project located in the southern tip of Greenland. The project has completed a Feasibility Study and is currently in the permitting phase. Last year was a busy time for the company as it completed a Feasibility Study, a mining licence application (draft submitted in December 2015) and pilot plant operations. Beneficiation pilot plant operations were completed at GTK in Finland in April 2015. This pilot plant treated approximately 30 tonnes of ore to producing almost 2 tonnes of rare earth mineral concentrate. Later in the year a hydrometallurgical pilot plant was performed which mimicked the Refinery process. This pilot plant was performed at Outotec’s Pori Research laboratories in Finland from September till October 2015. The pilot plant treated approximately 200 kilograms of concentrate over 4 split operating campaigns. Each campaign was performed to focus on the performance of a specific part of the refinery flowsheet. This allowed for full operating focus on a single unit operation to ensure that it was operating correctly. The pilot plant operations were quite successful with no major issues with the flowsheet identified through continuous operation. Some fine tuning of conditions was required to ensure adequate removal of impurities was performed with recycle streams incorporated. Overall the leach extractions observed in the pilot plant exceeded the design assumptions in the Feasibility Study. These programs were partially funded by the EURARE program. The EURARE program aims to encourage the sustainable development of European based rare earth projects. This has the goal of allowing Europe to become less reliant on importation of these key raw materials. The professionalism and performance of both GTK and Outotec contributed significantly to the success of the pilot plant

  16. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  17. Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

    2004-01-01

    L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

  18. Trend of field data on pipe wall thinning for BWR power plants

    International Nuclear Information System (INIS)

    Hakii, Junichi; Hiranuma, Naoki; Hidaka, Akitaka

    2009-01-01

    Strongly motivated by every stakeholder not to repeat Mihama Nuclear Power Station pipe rupture accident in August 2004, JSME Main Committee on Codes and Standards on Power Generation Facilities immediately launched a special task force to develop Rules on Pipe Wall Thinning Management for BWR, PWR and fossil Power Plants respectively. The authors describes the process of the development of Rules for BWR Power Plans from the view point of collections and analysis of fields data of pipe wall thinning. Through its activities, the authors confirmed the existing findings, like the effect of Oxygen injection, turbulence and dependence on coolant temperature, derived from series of laboratory-scaled experiments in FAC and coolant velocities effects in LDI. Further based upon the said proven findings with field data, they explain the adequacy of major concept of the rule such as separate treatment of FAC (Flow Accelerated Corrosion) and LDI (Liquid Droplet Impingement). (author)

  19. Waste Isolation Pilot Plant Overview

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Douglas James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-27

    The mission of Waste Isolation Pilot Plant (WIPP) is to demonstrate the safe, environmentally sound, cost effective, permanent disposal of Transuranic (TRU) waste left from production of nuclear weapons.

  20. WIPP: Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1984-01-01

    The following aspects of the Waste Isolation Pilot Plant are discussed briefly: history and site selection; salt as a disposal medium; transporting waste materials; early key events; impacts on New Mexico; project organization; and site certification profile

  1. New plant improves radwaste processing at the Tokai-2 BWR

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    New plant for radiowaste processing at the Tokaj-2 NPP, put in operation in September, 1986, is described. The plant includes five systems providing processing of drianage water, solid waste combustion, decrease of volume and solidification of concentrated wastes, waste storage and flushing water processing. Pressed tablets represent the final product of the waste processing. New plant enables to reduce sufficiently the volume of radioactive wastes

  2. An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

    International Nuclear Information System (INIS)

    Ting, K.; Chen, K.T.; Li, Y.C.; Hwang, S.H.; Chien, F.T.; Kang, J.C.

    2008-01-01

    Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs' and AOVs' safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee

  3. New technology for BWR power plant control and instrumentation

    International Nuclear Information System (INIS)

    Takano, Yoshiyuki; Nakamura, Makoto; Murata, Fumio.

    1992-01-01

    Nuclear power plants are facing strong demands for higher reliability and cost-performance in their control and instrumentation systems. To meet these needs, Hitachi is developing advanced control and instrumentation technology by rationalizing the conventional technology in that field. The rationalization is done through the utilization of reliable digital technology and optical transmission technology, and others, which are now commonly used in computer applications. The goal of the development work is to ensure safe, stable operation of the plant facilities and to secure harmony between man and machine. To alleviate the burdens of the operators, the latest electronic devices are being employed to create an advanced man-machine interface, and to promote automatic operation of the plant based upon the automatic operation of individual systems. In addition, the control and instrumentation system, including the safety system, incorporates more and more digital components in order to further enhance the reliability and maintainability of the plant. (author)

  4. Quantitative evaluation for training results of nuclear plant operator on BWR simulator

    International Nuclear Information System (INIS)

    Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

    1985-01-01

    Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

  5. Recent development of advanced BWR technology for plant application

    International Nuclear Information System (INIS)

    Horiuchi, Tetsuo; Sakurai, Mikio; Mase, Noriaki; Oyamada, Osamu; Nakadaira, Shiro.

    1988-01-01

    The development of advanced BWRs (ABWR) was completed in 1985. Through the authorization as the third improved and standardized plants of LWRs by the Ministry of International Trade and Industry, the detailed design phase of the actual project has begun, and the improved technology to be applied to the plants has been steadily verified and put in practical use. In the ABWRs, the operational capability, safety and economical efficiency as the general characteristics of the plants were further heightened by simplifying, heightening the performance and compactifying. Particularly internal pumps brought about the improvement of the operational capability and safety of the plants together with improved control rod driving system, besides promoting the simplification and compactification of the reactor system. Also the reinforced concrete containment vessels constructed into one body with the buildings have the compact structure and the solid property sufficiently withstanding hypothetical accidents, and contribute to the improvement of the economical efficiency and safety. These key improvement technologies completed their tests for practical use, and it was shown that the expected objectives are realized with the characteristics of high level, thus the steady steps toward the construction of the actual plants are promoted. (Kako, I.)

  6. Recent control and instrumentation systems for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Fujii, Hiroaki; Higashikawa, Yuichi; Sato, Hideyuki

    1990-01-01

    For the needs of the more stable operation of nuclear power stations, the upgrading of the measurement and control system for BWRs has been promoted by positively introducing remarkably advancing electronic technology. Further, it is aimed at to construct the synthetic digitized measurement and control system for nuclear power stations to heighten the operation reliability in ABWRs. As the first step of the development in the synthetic digitization, the monitoring and control system for radioactive waste treatment was put in practical use for No.5 plant of Kashiwazaki, Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. Hitachi Ltd. has promoted the development and the application to actual plants of the measurement and control system for BWRs, in which digital control technology, optical information transmission technology and the operation-supporting technology using a computer were utilized. Hereafter, it is intended to expand the application of digital measurement and control aiming at improving the reliability, operation performance and maintainability. The nuclear power plant control complex with advanced man-machine interface-90 (NUCAMM-90) was developed, and its application to actual plants is planned. (K.I.)

  7. Completion of high-efficiency BWR turbine plant 'Hamaoka unit No. 4'

    International Nuclear Information System (INIS)

    Tsuji, Kunio; Hamaura, Norikazu; Shibashita, Naoaki; Kazama, Seiichi

    1995-01-01

    Accompanying the increase of capacity of nuclear power plants in Japan, the plants having heightened economical efficiency, which are supported by the improvement of thermal efficiency and the reduction of dose, are demanded. Hitachi Ltd. has completed No. 4 turbine unit of 1137 MW output in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., which is the largest capacity machine in Japanese BWR plants. In this unit, the moisture separator heater, the steam turbine with high efficiency, and the hollow thread film condensate filter which treats the total flow rate of condensate are used as the reheating type BWR plant for the first time in Japan, and the plan of heightened economy and operation was adopted. It was confirmed by the trial for about 10 months that the planned performance was sufficiently satisfied, and the commercial operation was started in September, 1993. The features of the 1137 MW turbine unit are explained. The turbine is of tandem six-flow exhaust condensation type. Diffuser type low pressure turbine exhaust chambers, butterfly type combination intermediate valve are adopted. The stages with the blades having moisture-separating grooves were corrected. The reliability of the shaft system was improved. The adoption of the moisture separator heater and the application of the hollow thread film type condensate filter are explained. (K.I.)

  8. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  9. Operational experience of human-friendly control and instrumentation systems for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Makino, M.; Watanabe, T.; Suto, O.; Asahi, R.

    1987-01-01

    In recent BWR nuclear power plants in Japan, an advanced centralized monitoring and control system PODIA (Plant Operation by Displayed Information and Automation), which incorporates many operator aid functions, has been in operation since 1985. Main functions of the PODIA system as a computerized operator aid system are as follows. CRT displays for plant monitoring. Automatic controls and operation guides for plant operation. Stand-by status monitoring for engineered safety features during normal operation. Surveillance test procedure guides for engineered safety features. Integrated alarm display. The effectiveness of these functions have been proved through test and commercial operation. It has been obtained that operators have preferred PODIA much more than conventional monitoring and control systems

  10. Laguna Verde nuclear power plant: an experience to consider in advanced BWR design

    International Nuclear Information System (INIS)

    Fuentes Marquez, L.

    2001-01-01

    Laguna Verde is a BWR 5 containment Mark II. Designed by GE, two external re-circulation loops, each of them having two speed re-circulation pump and a flow control valve to define the drive flow and consequently the total core flow an power control by total core flow. Laguna Verde Design and operational experience has shown some insights to be considering in design for advanced BRW reactors in order to improve the potential of nuclear power plants. NSSS and Balance of plant design, codes used to perform nuclear core design, margins derived from engineering judgment, at the time Laguna Verde designed and constructed had conducted to have a plant with an operational license, generating with a very good performance and availability. Nevertheless, some design characteristics and operational experience have shown that potential improvements or areas of opportunity shall be focused in the advanced BWR design. Computer codes used to design the nuclear core have been evolved relatively fast. The computers are faster and powerful than those used during the design process, also instrumentation and control are becoming part of this amazing technical evolution in the industry. The Laguna Verde experience is the subject to share in this paper. (author)

  11. Development of remote automatic equipment for BWR power plants

    International Nuclear Information System (INIS)

    Sasaki, Masayoshi

    1984-01-01

    The development of remote control, automatic equipment for nuclear power stations has been promoted to raise the rate of operation of plants by shortening regular inspection period, to improve the safety and reliability of inspection and maintenance works by mechanization, to reduce the radiation exposure dose of workers and to reduce the manpower required for works. The taking-off of control rod drives from reactors and fixing again have been mechanized, but the disassembling, cleaning, inspection and assembling of control rod drives are manually carried out. Therefore, Hitachi Ltd. has exerted effort to develop the automatic equipment for this purpose. The target of development, investigation, the construction and function of the equipment, the performance and the effect of adopting it are reported. The equipment for the volume reduction of spent fuel channel boxes and spent control rods is developed since these are major high level radioactive solid wastes, and their apparent volume is large. Also the target of development, investigated things, the construction and function of the equipment, the performance and the effect of adopting it are reported. (Kako, I.)

  12. Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

    1995-01-01

    This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes

  13. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Cruz G, M.; Amador C, C.

    2013-10-01

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  14. Distributed control and data processing system with a centralized database for a BWR power plant

    International Nuclear Information System (INIS)

    Fujii, K.; Neda, T.; Kawamura, A.; Monta, K.; Satoh, K.

    1980-01-01

    Recent digital techniques based on changes in electronics and computer technologies have realized a very wide scale of computer application to BWR Power Plant control and instrumentation. Multifarious computers, from micro to mega, are introduced separately. And to get better control and instrumentation system performance, hierarchical computer complex system architecture has been developed. This paper addresses the hierarchical computer complex system architecture which enables more efficient introduction of computer systems to a Nuclear Power Plant. Distributed control and processing systems, which are the components of the hierarchical computer complex, are described in some detail, and the database for the hierarchical computer complex is also discussed. The hierarchical computer complex system has been developed and is now in the detailed design stage for actual power plant application. (auth)

  15. Consideration of severe accident issues for the General Electric BWR standard plant: Chapter 10

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982, the U.S. Nuclear Regulatory Commission (NRC) proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. GE provided appendices to the licensing documentation of its standard plant design, GESSAR II, which address severe accidents for the GE BWR/6 Mark III 238 nuclear island design. The GE submittals discuss the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at Three Mile Island (TMI), is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  16. On-line prediction of BWR transients in support of plant operation and safety analyses

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1983-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feed-water train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, and for coolant temperature. Control systems and trip logic are simulated for the nuclear steam supply system

  17. Improvement technique of sensitized HAZ by GTAW cladding applied to a BWR power plant

    International Nuclear Information System (INIS)

    Tujimura, Hiroshi; Tamai, Yasumasa; Furukawa, Hideyasu; Kurosawa, Kouichi; Chiba, Isao; Nomura, Keiichi.

    1995-01-01

    A SCC(Stress Corrosion Cracking)-resistant technique, in which the sleeve installed by expansion is melted by GTAW process without filler metal with outside water cooling, was developed. The technique was applied to ICM (In-Core Monitor) housings of a BWR power plant in 1993. The ICM housings of which materials are type 304 Stainless Steels are sensitized with high tensile residual stresses by welding to the RPV (Reactor Pressure Vessel). As the result, ICM housings have potential of SCC initiation. Therefore, the improvement technique resistant to SCC was needed. The technique can improve chemical composition of the housing inside and residual stresses of the housing outside at the same time. Sensitization of the housing inner surface area is eliminated by replacing low-carbon with proper-ferrite microstructure clad. High tensile residual stresses of housing outside surface area is improved into compressive side. Compressive stresses of outside surface are induced by thermal stresses which are caused by inside cladding with outside water cooling. The clad is required to be low-carbon metal with proper ferrite and not to have the new sensitized HAZ (Heat Affected Zone) on the surface by cladding. The effect of the technique was qualified by SCC test, chemical composition check, ferrite content measurement and residual stresses measurement etc. All equipment for remote application were developed and qualified, too. The technique was successfully applied to a BWR plant after sufficient training

  18. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Smith, S.K.; Lehnert, D.F.; Locke, R.K.

    1991-01-01

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  19. Recent operating experience during startup testing at latest 1100 MWe BWR-5 nuclear power plants

    International Nuclear Information System (INIS)

    Tanabe, Akira; Tateishi, Mizuo; Kajikawa, Makoto; Hayase, Yuichi.

    1986-01-01

    In June and September 1985, the latest two 1100 Mwe BWR-5 nuclear power plants started commercial operation about ten days earlier than initially expected without any unscheduled shutdown. These latest plants, 2F-3 and K-1, are characterized by an improved core with new 8 x 8 fuel assemblies, highly reliable control systems, advanced control room system and turbine steam full bypass system for full load rejection (2F3). This paper describes the following operating experiences gained during their startup testing. 1) Continuous operation at full load rejection. 2) Stable operation at natural circulating flow condition. 3) 31 and 23 hour short time start up operation. 4) 100-75-100 %, 1-8-1-14 hours daily load following operation. (author)

  20. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  1. Applicability study on a ceramic filter with hot-test conducted in a BWR plant

    International Nuclear Information System (INIS)

    Yamada, K.; Shirai, T.; Wada, M.; Nakamizo, H.

    1991-01-01

    Radioactive crud removal and filtration performance recovery by backwashing were examined with a BWR plant pool water using a ceramic filter element, 0.1 micron in nominal pore size and 0.2m 2 in filtration area. Totally 1114 hours filter operation were accumulated. Ten backwashings were accomplished during the test period. The following results were obtained. (1) Radioactive crud concentration in the filter effluent remained below 10 5 Bq/m 3 . (2) Both pressure loss through the filter and dose rate at the filter vessel surface were recovered to the initial level by each backwashing. The surface dose rate after backwashing was approximately 0.01mSv/h. According to these test results, it is confirmed that the ceramic filter is appropriate for the treatment of highly crud concentrated radioactive liquid, which is generated in nuclear facilities, such as spent fuel reprocessing plants. (author)

  2. Parameter identification of a BWR nuclear power plant model for use in optimal control

    International Nuclear Information System (INIS)

    Volf, K.

    1976-02-01

    The problem being considered is the modeling of a nuclear power plant for the development of an optimal control system of the plant. Current system identification concepts, combining input/output information with a-priori structural information are employed. Two of the known parameter identification methods i.e., a least squares method and a maximum likelihood technique, are studied as ways of parameter identification from measurement data. A low order state variable stochastic model of a BWR nuclear power plant is presented as an application of this approach. The model consists of a deterministic and a noise part. The deterministic part is formed by simplified modeling of the major plant dynamic phenomena. The moise part models the effects of input random disturbances to the deterministic part and additive measurement noise. Most of the model parameters are assumed to be initially unknown. They are identified using measurement data records. A detailed high order digital computer simulation is used to simulate plant dynamic behaviour since it is not conceivable for experimentation of this kind to be performed on the real nuclear power plant. The identification task consists in adapting the performance of the simple model to the data acquired from this plant simulation ensuring the applicability of the techniques to measurement data acquired directly from the plant. (orig.) [de

  3. Application of passive auto catalytic recombiner (PAR) for BWR plants in Japan

    International Nuclear Information System (INIS)

    Kobayashi, K.; Murano, K.; Yamanari, S.; Yamamoto, Y.

    2001-01-01

    The passive auto-catalytic recombiner (PAR), which can recombine flammable gases such as hydrogen and oxygen with each other to avoid an explosion in case of a loss-of-coolant accident (LOCA), installed in the primary containment vessel does not require a power supply or dynamic equipment, while the existing flammability gas control system (FCS) of most BWRs as an outer loop of the primary containment vessel needs them to make flammable gases circulate through blowers and heaters in the system. PAR offers a number of advantages over existing FCS, such as high reliability, low cost due to much smaller amount of materials needed, good maintainability, good operability in case of a LOCA, and smaller space for installation. An experimental study has been carried out for the purpose of solving the problems of applying PAR to Japanese BWR plants instead of existing FCS, in which we grasped the basic characteristics of PAR. (author)

  4. Computerized operator support system with new man-machine interface for BWR power plants

    International Nuclear Information System (INIS)

    Monta, K.; Naito, N.; Sugawara, M.; Sato, N.; Mori, N.; Tai, I.; Fukumoto, A.; Tsuchida, M.

    1984-01-01

    Improvement of the man-machine interface of nuclear power plants is an important contribution to the further enhancement of operational safety. In addition, recent advances in computer technology seem to offer the greatest opportunity to date for achieving improvement in the man-machine interface. The development of a computerized operator support system for BWRs has been undertaken since 1980 with the support of the Japanese Government. The conceptual design of this system is based on the role of the operators. The main functions are standby system management, disturbance analysis and post-trip operational guidance. The objective of the standby system management is to monitor the standby status of the engineered safety feature during normal operation to assure its proper functioning at the onset of emergency situations. The disturbance analysis system detects disturbances in the plant in their early stages and informs the plant operators about, for example, the cause of the disturbances, the plant status and possible propagations. Consequently, operators can take corrective actions to prevent unnecessary plant shutdown. The objective of the post trip operational guide is to support operators in diagnosis and corrective action after a plant trip. Its functions are to monitor the performance of the engineered safety feature, to identify the plant status and to guide the appropriate corrective action to achieve safe plant shutdown. The information from the computerized operator support system is supplied to operators through a colour CRT operator console. The authors have evaluated the performance of various new man-machine interfacing tools and proposed a new operator console design. A prototype system has been developed and verification/validation is proceeding with a BWR plant simulator. (author)

  5. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  6. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  7. Cooperative control scheme for an HVDC system connected to an isolated BWR nuclear power plant

    International Nuclear Information System (INIS)

    Sakurai, T.; Goto, K.; Kawai, T.; Matori, I.; Nakao, T.; Watanabe, A.

    1983-01-01

    This paper describes a cooperative control system to achieve stable operation of an isolated BWR nuclear plant linked to an HVDC system. In the proposed control system, under normal conditions the power plant is controlled according to the generating power reference and the generator frequency deviation is adjusted by converter power control. Such frequency control is also effective in the case of AC-DC system faults. In addition to the frequency control, an overload control is provided with the HVDC system, where the DC transmission power in the sound poles is increased due to a fault detection signal from the faulty pole. Effects of the above mentioned control systems were studied using digital dynamic programs. The sets of simulation results confirmed that in the case of a DC single pole fault, the plant is able to continue operation without any use of the turbine speed control units even for a restarting failure in the faulty pole. In case of a DC two pole fault, the plant is able to continue operation, being assisted by turbine speed control units when restarting in the faulty poles succeeds. In case of an AC three-line to ground fault near the AC terminal of the converter at the sending or receiving end, the system is able to continue stable operation, being supplemented by the turbine control unit when the faulty section of the AC system is isolated by a main or back-up relaying system

  8. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    Eckardt, B.

    1991-01-01

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO 2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  9. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng [Atomic Energy Council, Taoyuan City, Taiwan (China). Inst. of Nuclear Energy Research

    2017-11-15

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  10. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    International Nuclear Information System (INIS)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng

    2017-01-01

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  11. ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics

    International Nuclear Information System (INIS)

    Fuller, L.C.; Myers, M.L.

    1975-01-01

    1 - Description of problem or function: ORCOST2 estimates the cost of electrical energy production from single-unit steam-electric power plants. Capital costs and operating and maintenance costs are calculated using base cost models which are included in the program for each of the following types of plants: PWR, BWR, HTGR, coal, oil, and gas. The user may select one of several input/output options for calculation of capital cost, operating and maintenance cost, levelized energy costs, fixed charge rate, annual cash flows, cumulative cash flows, and cumulative discounted cash flows. Options include the input of capital cost and/or fixed charge rate to override the normal calculations. Transmission and distribution costs are not included. Fuel costs must be input by the user. 2 - Method of solution: The code follows the guidelines of AEC Report NUS-531. A base capital-cost model and a base operating- and maintenance-cost model are selected and adjusted for desired size, location, date, etc. Costs are discounted to the year of first commercial operation and levelized to provide annual cost of electric power generation. 3 - Restrictions on the complexity of the problem: The capital cost models are of doubtful validity outside the 500 to 1500 MW(e) range

  12. A nonlinear 3D real-time model for simulation of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.

    1982-02-01

    A nonlinear transient model for BWR nuclear power plants which consists of a 3D-core (subdivided into a number of superboxes, and with parallel flow and subcooled boiling), a top plenum, steam removal and feed water systems and main coolant recirculation pumps is given. The model describes the local core and global plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. The model which forms the basis for the digital code GARLIC-B (Er et al. 82) is aimed to be used on an on-site process computer in parallel to the actual reactor process (or even in predictive mode). Thus, special measures had to be taken into account in order to increase the computational speed and reduce the necessary computer storage. This could be achieved by - separating the neutron and power kinetics from the xenon-iodine dynamics, - treating the neutron kinetics and most of the thermodynamics and hydrodynamics in a pseudostationary way, - developing a special coupling coefficient concept to describe the neutron diffusion, calculating the coupling coefficients from a basic neutron kinetics code, - combining coarse mesh elements into superboxes, taking advantage of the symmetry properties of the core and - applying a sparse matrix technique for solving the resulting algebraic power equation system. (orig.) [de

  13. Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

    International Nuclear Information System (INIS)

    D'Ardenne, W.H.; Sherwood, G.G.

    1985-01-01

    The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

  14. Global vibrations in the wetwell condensation process caused by LOCA in BWR plants

    International Nuclear Information System (INIS)

    Bjoerndahl, O.; Andersson, Magnus

    1998-12-01

    During the last years a substantial part of third part review work related to dynamical loadings has been review of loading specifications dealing with vibrations in containment building related to so called LOCA-events in Swedish BWR plants. Compared to other loading categories characterised as global vibrations these secondary effects of LOCA-events are complex to analyse. One experience from the review work at SAQ up to now is that it is not fully clear what prediction methods and what model idealisations are the most adequate for structural integrity verification on mechanical systems as pressure vessels and piping under such loading conditions. At SAQ Teknik a project work has been carried out to investigate the general status of the methodology used today in Sweden and a work to in the long term develop simplified prediction models and methods for the loading categories condensation oscillations (CO) and chugging (CH). The work was initially concentrated on a study of the background of the methodology which was developed for these type of loading in American BWR-containments of the Mark-II design. The methodology was developed by General Electric, GE, in cooperation with the Mark-II plant owners. The methodology used in Sweden to predict vibrations in BWR containments of this design is with some minor modifications very close to technique developed by GE. The methodology developed by GE is the only accepted by USNRC for the Mark-II design and could be found as reference in Standard Review Plan 6.2.1.1.C, Rev 6 - August 1984. Based on identical physical assumptions about the dynamic behaviour of the building structure and the water in the suppression pool, mathematical models are derived in this report for predictions of secondary structure response spectra for loading conditions as global vibrations during CO and CH. Based on parameters identified by so called one pipe experiments responses my be predicted. By use of these derived mathematical models as a

  15. ASTM standards associated with PWR and BWR power plant licensing, operation and surveillance

    International Nuclear Information System (INIS)

    McElroy, W.N.; McElroy, R.J.; Gold, R.; Lippincott, E.P.; Lowe, A.L. Jr.

    1994-01-01

    This paper considers ASTM Standards that are available, under revision, and are being considered in support of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) Nuclear Power Plant (NPP) licensing, regulation, operation, surveillance and life attainment. The current activities of ASTM Committee E10 and its Subcommittees E10.02 and current activities of ASTM Committee E10 and its Subcommittees E10.02 and E10.05 and their Task Groups (TG) are described. A very important aspect of these efforts is the preparation, revision, and balloting of standards identified in the ASTM E706 Standard on Master Matrix for Light Water Reactor (LWR) Pressure Vessel (PV) Surveillance Standards. The current version (E706-87) of the Master Matrix identifies 21 ASTM LWR physics-dosimetry-metallurgy standards for Reactor Pressure Vessel (RPV) and Support Structure (SS) surveillance programs, whereas, for the next revision 34 standards are identified. The need for national and international coordination of Standards Technology Development, Transfer and Training (STDTT) is considered in this and other Symposium papers that address specific standards related physics-dosimetry-metallurgy issues. 69 refs

  16. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    International Nuclear Information System (INIS)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu

    1999-01-01

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

  17. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

    International Nuclear Information System (INIS)

    Escorcia O, D.; Salazar S, E.

    2016-09-01

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  18. Experiences with monitoring and control of microbiological growth in the standby service water system of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Zisson, P.S.; Whitaker, J.M.; Neilson, H.L.; Mayne, L.L.

    1995-01-01

    In 1989, the Unites States Nuclear Regulatory Commission formally recognized the potential for nuclear accidents resulting from microbiological causes. Such causes range from loss of heat transfer due to microbiological fouling, to loss of system integrity caused by microbiologically influenced corrosion (MIC). As a result of these potential problems, monitoring, mitigation, and control procedures must be developed by all regulated plants. In developing a control and mitigation strategy for the standby service water system of a boiling water reactor (BWR) nuclear power plant, numerous monitoring techniques were employed to evaluate effectiveness. This paper describes the monitoring techniques that were evaluated, and those that ultimately proved to be effective

  19. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Johnsen, G.W.; Lellouche, G.S.

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  20. Development and Evaluation of cooperative control system for an HVDC transmission system connected with an isolated BWR power plant

    International Nuclear Information System (INIS)

    Horiuchi, Susumu; Hara, Tsukusi; Matori, Iwao; Hirayama, Kaiichirou.

    1987-01-01

    In the cooperative control system developed for an HVDC transmission system connected with an isolated BWR power plant, the equilibrium state between power plant output and DC transmission power is examined by way of the detection of the generator frequency. And, thereby start-up and shutdown of the DC system and controlling of the transmission power are made, so that the signal transmission with the power plant becomes unnecessary, enabling the easy cooperative operation. In order to investigate validity of this control system, various digital simulation and simulator test with the control system were carried out. In this way, behavior of the power plant and stability of the DC transmission system were evaluated in the connection to the DC system at power plant start-up, follow of the transmission power in change of the power plant output and in various system failures. (Mori, K.)

  1. Extractive metalurgical pilot plant. Project and installation

    International Nuclear Information System (INIS)

    Paula, H.C.B.; Rolim, T.L.; Santana, A.O. de; Santos, F.S.M. dos; Dantas, C.C.

    1986-01-01

    An extractive metalurgical pilot plant with a flow capacity of 200l/h of phosphoric leach, recovering 80% of the uranium content has been designed and installed. Starting from the diagrams of the chemical process in the laboratory scale, the equipment worksheet of the basic project were developed. The procedure for dimensioning and positioning of each component is described. An isometric figure and the pilot plant lay-out are included. The pilot plant occupying 41 m 2 has been tested and operates at its nominal capacity. (author) [pt

  2. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  3. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  4. Corrosion issues in the BWR and their mitigation for plant life extension

    International Nuclear Information System (INIS)

    Gordon, B.M.

    1988-01-01

    Corrosion is a major service life limiting mechanism for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the BWR, stress corrosion cracking of piping has been the major source of concern where extensive research has led to a number of qualified remedies and currently > 90% of susceptible welds have been mitigated or replaced. Stress corrosion cracking of reactor internals due to the interaction of irradiation, as discussed elsewhere in this conference, is also a possible life limiting phenomenon. This paper focusses on two corrosion phenomena in the BWR which have only recently been identified as impacting the universal goal of BWR life extension: the general corrosion of containment structures and the erosion-corrosion of carbon steel piping

  5. Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Yuji, Nemoto; Toshihisa, Tsukiyama; Shigeki, Nemezawa [Hitachi. Ltd., Saiwai-cho, Hitachi (Japan); Tadashi, Yamasaki; Hidetsugu, Okada [Chubu Electric Power Company, Inc., Odaka-cho, Midori-ku Nagoya (Japan)

    2007-07-01

    The Monte Carlo N-Particle transport code (MCNP) was adopted to calculate the skyshine dose from the turbine building of a BWR plant for obtaining precise estimations at the site boundary. In MCNP calculation, the equipment and piping arranged on the operating floor of the turbine building were considered and modeled in detail. The inner and outer walls of the turbine building, the shielding materials around the high-pressure turbine, and the piping connected from the moisture separator to the low-pressure turbine were all considered. A three-step study was conducted to estimate the applicability of MCNP code. The first step is confirming the propriety of calculation models. The atmospheric relief diaphragms, which are installed on top of the low-pressure turbine exhaust hood, are not considered in the calculation model. There was little difference between the skyshine dose distributions that were considered when using and not using the atmospheric relief diaphragms. The calculated dose rates agreed well with the measurements taken around the turbine. The second step is estimating the dose rates on the outer roof surface of the turbine building. This calculation was made to confirm the dose distribution of gamma-rays on the turbine roof before being scattered into the air. The calculated dose rates agreed well with the measured data. The third step is making a final confirmation by comparing the calculations and measurements of skyshine dose rates around the turbine building. The source terms of the main steam system are based on the measured activity data of N-16 and C-15. As a conclusion, we were able to calculate reasonable skyshine dose rates by using MCNP code. (authors)

  6. Design of a uranium recovery pilot plant

    International Nuclear Information System (INIS)

    1984-01-01

    The engineering design of a pilot plant of uranium recover, is presented. The diagrams and specifications of the equipments such as pipelines, pumps, values tanks, filters, engines, etc... as well as metallic structure and architetonic design is also presented. (author)

  7. Plant analyzer development for high-speed interactive simulation of BWR plant transients

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.

    1986-01-01

    Advanced modeling techniques have been combined with modern, special-purpose peripheral minicomputer technology to develop a plant analyzer which provides realistic and accurate predictions of plant transients and severe off-normal events in nuclear power plants through on-line simulations at speeds of approximately 10 times faster than actual process speeds. The new simulation technology serves not only for carrying out routinely and efficiently safety analyses, optimizations of emergency procedures and design changes, parametric studies for obtaining safety margins and for generic training but also for assisting plant operations. Five modeling principles are presented which serve to achieve high-speed simulation of neutron kinetics, thermal conduction, nonhomogeneous and nonequilibrium two-phase flow coolant dynamics, steam line acoustical effects, and the dynamics of the balance of plant and containment systems, control systems and plant protection systems. 21 refs

  8. Plant analyzer for high-speed interactive simulation of BWR plant transients

    International Nuclear Information System (INIS)

    Cheng, H.S.; Lekach, S.V.; Mallen, A.N.; Wulff, W.; Cerbone, R.J.

    1984-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology was utilized to develop a plant analyzer which affords realistic predictions of plant transients and severe off-normal events in LWR power plants through on-line simulations at speeds up to 10 times faster than actual process speeds. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the entire balance of the plant. Reactor core models include point kinetics with reactivity feedback due to void fraction, fuel temperature, coolant temperature, and boron concentration as well as a conduction model for predicting fuel and clad temperatures. Control systems and trip logic for plant protection systems are also simulated. The AD10 of Applied Dynamics International, a special-purpose peripheral processor, is used as the principal hardware of the plant analyzer

  9. Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

    2014-07-01

    After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

  10. Basic study on BWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Jyohko, Shingo; Dohgo, Hirofumi

    2015-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the severe accident that LOCA occurs but ECCS fails the water irrigation into the reactor core. The simulation experiments were carried out for the cases that LOCA has occurred in the main steam piping or in the recirculation piping, respectively. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for LOCA area of up to 100 cm"2 for both type LOCA. In the case of main steam system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the recirculation system LOCA of 2000 cm"2 area, a drop of the core water level was extremely in comparison with main steam system LOCA, and the fuel assemblies were completely exposed during up to 30 min, to the irrigation from approximately 100 sec, after LOCA occurrence. Therefore, the fuel temperature during the irrigation had reached approximately 1900degC. Thus, the fuel cladding were damaged approximately less than 10%, and H_2 concentration in the containment vessel was approximately 9% which did not exceed H_2 detonation limit of 13% but exceeded H_2 explosion limit of 4%. However, the containment vessel internal pressure was settled around design pressure value of containment vessel. As the results, some core damage could not be avoided, but soundness of the containment vessel, which should take the role of 'confine', was found to be secured. (author)

  11. Basic study on BWR plant behavior under the condition of severe accident (2)

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Ueda, Masataka; Sasaki, Hajime

    2016-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the two types of severe accidents that LOCA occurs but ECCS fails the water irrigation into the reactor core and SBO occurs and at the same time the reclosed failure of SRV occurs. The simulation experiments were carried out for the cases that LOCA has occurred in the main feed-water piping. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for small and middle LOCA area. In the case of main feed-water system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the accident of SBO with reclosed failure of SRV, it has been shown that the accidents continue to progress rapidly as compared with the case of normally operating of SRV. Because SRV has the function that keep the inside pressure of reactor core by repeating opened and closed in response of the inside pressure and prevent the decrease of water level inside reactor core. However, if the irrigation into the reactor core was carried out 30 min after SBO occurrence, the core had little damage and also the H_2 concentration in the containment vessel did not exceed limits of H_2 explosion. Further, as for the accident of reclosed failure of SRV, it has been shown that there are very good correspondence with the simulation results of main steam piping LOCA of area 180 cm"2 corresponding to the inlet cross-sectional area SRV installed on the piping

  12. Direct torus venting analysis for Chinshan BWR-4 plant with MARK-I containment

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

    Highlights: • Study the effectiveness of Direct Torus Venting System (DTVS) during extended SBO of 24 h for Chinshan MARK-I plant. • Containment response is analyzed by GOTHIC based on boundary conditions from RETRAN calculation. • Analyses are performed with and without DTVS, respectively. • Suppression pool is sub-divided and thermal stratification is observed. - Abstract: The Chinshan plant, owned by Taiwan Power Company, has twin units of BWR-4 reactor and MARK-I containment. Both units have been operating at rated core thermal power of 1840 MWt. The existing Direct Torus Venting System (DTVS) is the main system used for venting the containment during the extended station blackout event. The purpose of this paper is to study the effects of the DTVS venting on the response of the containment pressure and temperature. The reactor is depressurized by manually opening the safety relief valves (SRVs) during the SBO, which causes the mass and energy to be discharged into and heat up the suppression pool. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The DTVS model is established in the GOTHIC model based on the venting size, venting piping loss, venting initiation time, and venting source. The lumped volume model, 1-D coarse-mesh model, and 3-D coarse-mesh model are considered in the torus volume. The calculation is first done without DTVS venting to establish a reference basis. Then a case with DTVS available is performed. Comparison of the two cases shows that the existing DTVS design is effective in mitigating the severity of the containment pressure and temperature transients. The results also show that the 1-D coarse-mesh model may not be appropriate since a

  13. Application of Continuous and Structural ARMA modeling for noise analysis of a BWR coupled core and plant instability event

    International Nuclear Information System (INIS)

    Demeshko, M.; Dokhane, A.; Washio, T.; Ferroukhi, H.; Kawahara, Y.; Aguirre, C.

    2015-01-01

    Highlights: • We demonstrate the first application of a novel CSARMA method. • We analyze the instability occurred in a Swiss BWR plant during power ascension. • Benchmarked the results against STP analysis. • The CSARMA results are consistent with the background physics and the STP results. • The instability was caused by disturbances in the pressure control system. - Abstract: This paper presents a first application of a novel Continuous and Structural Autoregressive Moving Average (CSARMA) modeling approach to BWR noise analysis. The CSARMA approach derives a unique representation of the system dynamics by more robust and reliable canonical models as basis for signal analysis in general and for reactor diagnostics in particular. In this paper, a stability event that occurred in a Swiss BWR plant during power ascension phase is analyzed as well as the time periods that preceded and followed the event. Focusing only on qualitative trends at this stage, the obtained results clearly indicate a different dynamical state during the unstable event compared to the two other stable periods. Also, they could be interpreted as pointing out a disturbance in the pressure control system as primary cause for the event. To benchmark these findings, the frequency-domain based signal transmission-path (STP) method is also applied. And with the STP method, we obtained similar relationships as mentioned above. This consistency between both methods can be considered as being a confirmation that the event was caused by a pressure control system disturbance and not induced by the core. Also, it is worth noting that the STP analysis failed to catch the relations among the processes during the stable periods, that were clearly indicated by the CSARMA method, since the last uses more precise models as basis

  14. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  15. Exxon Nuclear WREM-based NJP-BWR ECCS evaluation model and example application to the Oyster Creek Plant

    International Nuclear Information System (INIS)

    Krysinski, T.L.; Bjornard, T.A.; Steves, L.H.

    1975-01-01

    A proposed integrated ECCS model for non-jet pump boiling water reactors is presented, using the RELAP4-EM/BLOWDOWN and RELAP4-EM/SMALL BREAK portions of the Exxon Nuclear WREM-based Generic PWR Evaluation Model coupled with the ENC NJP-BWR Fuel Heatup Model. The results of the application of the proposed model to Oyster Creek are summarized. The results of the break size sensitivity study using the proposed model for the Oyster Creek Plant are presented. The application of the above results yielded the MAPLHGR curves. Included are a description of the proposed non-jet pump boiling water reaction evaluation model, justification of its conformance with TOCFR50, Appendix K, the adopted Oyster Creek plant model, and results of the analysis and sensitivity studies. (auth)

  16. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  17. Pilot plant for exploitation of geothermal waters

    Directory of Open Access Journals (Sweden)

    Stojiljković Dragan T.

    2006-01-01

    Full Text Available In Sijarinska spa, there are some 15 mineral and thermomineral springs, that are already being used for therapeutic purposes. For the exploitation of heat energy boring B-4 is very interesting. It is a boring of a closed type, with the water temperature of about 78°C and a flow rate of about 33 l/s. Waters with the flow rate of about 6 l/s are currently used for heating of the Gejzer hotel, and waters of the flow rate of about 0,121 l/s for the pilot drying plant. The paper presents this pilot plant. .

  18. Risk-informed decision-making analysis for the electrical raceway fire barrier systems on a BWR-4 plant

    International Nuclear Information System (INIS)

    Wu, Ching-Hui; Lin, Tsu-Jen; Kao, Tsu-Mu; Chen, Chyn-Rong

    2003-01-01

    This paper describes a risk-informed decision-making approach used to resolve the fire barrier issue in a BWR-4 nuclear plant where Appendix R separation requirements cannot be met without installing additional fire protection features such as electrical raceway fire barrier system. The related risk measures in CDF (core damage frequency) and LERF (large early release frequency) of the fire barrier issue can be determined by calculating the difference in plant risks between various alternative cases and that met the requirement of the Appendix R. In some alternative cases, additional early-detection and fast-response fire suppression systems are suggested. In some other cases, cable re-routing of some improper layout of non-safety related cables are required. Sets of fire scenarios are re-evaluated more detailed by reviewing the cable damage impact for the BWR-4 plant. The fire hazard model, COMPBRM III-e, is used in this study and the dominant results in risk measures are benchmarked with the CFD code, FDS 2.0, to ensure that the risk impact of fire barrier is estimated accurately in the risk-informed decision making. The traditional deterministic qualitative methods, such as defense-in-depth, safety margin and post-fire safety shutdown capability are also proceeded. The value-impact analysis for proposed alternatives of fire wrapping required by Appendix R has been completed for technical basis of the exemption on Appendix R application. The outcome of the above analysis should be in compliance with the regulatory guidelines (RG) 1.174 and 1.189 for the applications in the risk-informed decision-making of the fire wrapping issues. (author)

  19. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    International Nuclear Information System (INIS)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

  20. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    2000-10-01

    plant components addressed in the reports. This report addresses the metal components of BWR containment systems. The primary ageing mechanisms that may potentially impact the structural capacity, leaktight integrity, or service life of BWR containments are corrosion of metal components and stress corrosion cracking of bellows. Other potential degradation mechanisms include fatigue and mechanical wear. Areas of concern are where surfaces are inaccessible for inspection (e.g areas adjacent to floors, where the containment vessel is embedded in concrete, and locations adjacent to equipment or other structures)

  1. Characterization of solidified radioactive wastes produced at Montalto di Castro BWR plant with reference to the site storage

    International Nuclear Information System (INIS)

    Donato, A.; Ricci, G.; Pace, A.

    1985-01-01

    The cement solidification of the Montalto di Castro BWR plant radwastes has been studied both from the point of view of the mixtures of formulation and of the product characterization. Five radwaste types and mixtures of them have been taken into consideration, determining the best chemical formulations starting from the compressive strenght as leading parameter. The solidified products have been characterized from the point of view of the freeze and thawing resistance, the water immersion resistance, the leachability, the dimensional changes and the free standing water. All the tests have been performed taking into account the real site conditions, so the leaching tests and the water immersion tests have been carried out using sea water and table water as leachant

  2. Containment venting as a mitigation technique for BWR MARK I plant ATWS

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1987-01-01

    Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure

  3. 7 CFR 1412.48 - Planting Transferability Pilot Project.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Planting Transferability Pilot Project. 1412.48... and Peanuts 2008 through 2012 § 1412.48 Planting Transferability Pilot Project. (a) Notwithstanding § 1412.47, for each of the 2009 and subsequent crop years, the Planting Transferability Pilot Project...

  4. Standard Technical Specifications General Electric plants, BWR/4:Bases (Sections 3.4-3.10). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the specifications for all chapters and sections of the improved STS. Volume 2 contains he Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4-3.10 of the improved STS

  5. Standard Technical Specifications General Electric plants, BWR/4: Bases (Sections 2.0-3.3). Volume 2, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved ST or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume I contains the Specifications for all chapters and sections of the improved STS. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1-3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4-3.10 of the improved STS

  6. Fusion pilot plant scoping study

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Blevins, P.J.; Brunnader, H.; Natalizio, A.; Cumyn, P.; Dean, B.; Smith, S.; Galambos, J.; Holloway, C.; Stremlaw, J.; Williams, G.

    1994-05-01

    CFFTP Pilot is representative of a class of machines that, like NPD in the CANDU development program, could test the key reactor core technologies on an integrated power reactor relevant system (materials, conditions, configuration). But in order to reduce costs, the machine would operate at reduced neutron flux relative to a power reactor, would not produce electricity, and would not test superconducting magnets. This design shows research directions towards a machine that could provide integrated nuclear testing (but not ignition physics) at a cost of about 1/3 ITER CDA. The test volume - the outboard blanket volume - would be comparable to the test port volume on ITER CDA, while the fluence and power density would be about 1/4 ITER CDA. 91 refs., 43 tabs., 45 figs

  7. Fusion pilot plant scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Gierszewski, P J; Blevins, P J; Brunnader, H; Natalizio, A [Canadian Fusion Fuels Technology Project, Toronto, ON (Canada); Cumyn, P [Canatom Ltd., Montreal, PQ (Canada); Dean, B; Smith, S [Wardrop (W.L.) and Associates Ltd., Winnipeg, MB (Canada); Galambos, J [Oak Ridge National Lab., TN (United States); Holloway, C [Spar Aerospace Ltd., Toronto, ON (Canada); Stremlaw, J [Monenco AGRA Inc., Calgary, AB (Canada); Williams, G [Spectrum Engineering Corp., Peterborough, ON (Canada)

    1994-05-01

    CFFTP Pilot is representative of a class of machines that, like NPD in the CANDU development program, could test the key reactor core technologies on an integrated power reactor relevant system (materials, conditions, configuration). But in order to reduce costs, the machine would operate at reduced neutron flux relative to a power reactor, would not produce electricity, and would not test superconducting magnets. This design shows research directions towards a machine that could provide integrated nuclear testing (but not ignition physics) at a cost of about 1/3 ITER CDA. The test volume - the outboard blanket volume - would be comparable to the test port volume on ITER CDA, while the fluence and power density would be about 1/4 ITER CDA. 91 refs., 43 tabs., 45 figs.

  8. Experience for plant monitoring design in Italian BWR NPP and future trends in man-machine interface

    International Nuclear Information System (INIS)

    Maestri, F.; Sepielli, M.

    1987-01-01

    TMI accidental sequence and daily-gained operating experience on italian and abroad NPPs have affected in depth the approach to the design of information presentation to the Control Room staff. It has been cleared that most problems in plant operation arise from a poor and inadequate information system. The main lacks have been identified in the Control Room lay-out and information organization. This has pushed designers both to improve the Control Room environment and to better exploit the computer data processing and data presentation capabilities. The paper deals with the basic criteria for the design and the design review of a computerized system to be inserted in a hybrid Control Room in Italian 981 Mwe BWR-6 NPP, where the concepts outlined above were taken-up from the very beginning. The Control Room keeps conventional instrumentation arranged in a human-factor lay-out, according to post-TMI requirements, and adds a powerful computer-based information system for advanced alarm presentation and plant supervision during both normal and emergency conditions with high data reliability. Colour videounits and operating panels are functionally integrated to create powerful operator work-stations. Emphasis is mostly given on the revision work for video-unit displays and Man-System Communication carried out in cooperation with Halden Reactor Project human factor and plant operation experts. The work peculiarity has been a strong care on the integration between conventional and computerized information presentation, with particular regard to common information and code consistency. (author)

  9. Ion exchange/adsorbent pilot plant

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    A decontamination of greater than 99% of the actinides and fission products contained in radioactive waste water can be obtained using ion exchange resins. A system for achieving this result is described in this paper. This ion exchange pilot-plant design is the culmination of five years of study of the decontamination of radioactive waste streams by ion exchange resins and other adsorbents at Mound. In order to maintain maximum flexibility of treatments, this pilot-plant design is a conceptual design with specific flows, resins, and column specifications, but with many optional features and no rigid equipment specifications. This flexibility allows the system to be amenable to almost any radioactive waste stream. Very specific designs can be constructed from this conceptual design for the treatment of any specific waste stream. Operating and capital costs are also discussed. 1 figure, 5 tables

  10. Radioactive-waste isolation pilot plant

    International Nuclear Information System (INIS)

    Weart, W.D.

    1977-01-01

    The objective of the Waste Isolation Pilot Plant (WIPP) program is to demonstrate the suitability of bedded salt, specifically, the bedded salt deposits in the Los Medanos area of southeastern New Mexico, as a disposal medium for radioactive wastes. Our program responsibilities include site selection considerations, all aspects of design and development, technical guidance of facility operation, environmental impact assessment, and technical support to ERDA for developing public understanding of the facility

  11. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.; Hoeld, A.; Lupas, O.

    1982-04-01

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL) [de

  12. Comparison of the corrosion potential for stainless steel measured in-plant and in laboratory during BWR normal water chemistry conditions

    International Nuclear Information System (INIS)

    Molander, A.; Pein, K.; Tarkpea, P.; Takagi, Junichi; Karlberg, G.; Gott, K.

    1998-01-01

    To obtain reliable crack growth rate date for stainless steel in BWR environments careful laboratory simulation of the environmental conditions is necessary. In the plant the BWR normal water chemistry environment contains hydrogen peroxide, oxygen and hydrogen. However, in crack growth rate experiments in laboratories, the environment is normally simulated by adding 200 ppb oxygen to the high temperature water. Thus, as hydrogen peroxide is a more powerful oxidant than oxygen, it is to be expected that a lower corrosion potential will be measured in the laboratory than in the plant. To resolve this issue this work has been performed. In-plant and laboratory measurements have often been performed with somewhat different equipment, due to the special requirements concerning in-plant measurements. In this work such differences have been avoided and two identical sets of equipment for electrochemical measurements were built and used for measurements in-plant in a Swedish BWR and in high purity water in the laboratory. The host plant was Barsebaeck 1. Corrosion potential monitoring in-plant was performed under both NWC (Normal Water Chemistry) and HWC (Hydrogen Water Chemistry) conditions. This paper is, however, focused on NWC conditions. This is due to the fact, that the total crack growth obtained during a reactor cycle, can be determined by NWC conditions, even for plants running with HWC due to periodic stops in the hydrogen addition for turbine inspections or failure of the dosage or hydrogen production equipment. Thus, crack growth data for NWC is of great importance both for BWRs operating with HWC and NWC. Measurements in-plant and in the laboratory were performed during additions of oxygen and hydrogen peroxide to the autoclave systems. The corrosion potentials were compared for various conditions in the autoclaves, as well as versus in-plant in-pipe corrosion potentials. (J.P.N.)

  13. Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plant:2002

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Suzuki, Riichiro; Komura, Seiichi; Kudo, Yoshiro; Yamanaka, Akihiro; Oomizu, Satoru; Kitamura, Hideya; Nagata, Yoshifumi

    2003-01-01

    To secure fuel integrity, a Light Water Reactor (LWR) core is designed so that no boiling transition (BT) should take place in fuel assemblies and excessive rise in fuel cladding temperature due to deteriorated that transfer should be avoided in Anticipated Operational Occurrences (AOO). In some AOO in a Boiling Water Reactor (BWR), however, the rise in reactor power could be limited by SCRAM or void reactivity effect. Recent studies have provided accumulated knowledge that even if BT takes place in fuel assemblies, the rise in fuel cladding temperature could be so small that it will not threat to fuel integrity, as long as the BT condition terminates within a short period of time. In addition, appropriate methods have been developed to evaluate the cladding temperature during dryout. This standard provides requirements in the assessment of fuel integrity under AOO in which limited-BT condition is temporarily reached and the propriety of reusing a fuel assembly that has experienced limited-BT condition. The standard has been approved by the Atomic Energy Society of Japan following deliberation by impartial members for two and half years. It is now expected that this standard will provide an effective measure for the rational expansion of fuel design and operational margin. (author)

  14. Plan for the civil reprocessing pilot plant of China

    International Nuclear Information System (INIS)

    Wang, D.Y.; Chen, M.

    1987-01-01

    Based on the R and D work, experience on plant operation and site situation, the necessity and feasibility of building a pilot plant for civil reprocessing in China are discussed. The capacity of 100 kg HM/day (LWR) and 3 kg HM/day (MTR) has been proposed. The plant consists of cold testing facility and hot pilot facility. It is expected to complete the pilot plant in 1990's. This paper also describes the purpose, scale, process and equipment of the pilot plant

  15. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  16. A review on pilot plant development models

    International Nuclear Information System (INIS)

    Rosli Darmawan

    2005-01-01

    After more than 30 years, MINT has been able to produce many new findings, products and processes. Some of these have been able to penetrate local and international markets. This was achieved through a systematic commercialisation program practiced in MINT with its technological chain and MINT Technology Park program. This paper will review the development process of MINT pilot plants and compare them with a few other models from other institutions in Malaysia and abroad. The advantages and disadvantages of each model are reviewed and a discussion against MINT's model is presented. (Author)

  17. Vitrification pilot plant experiences at Fernald, Ohio

    International Nuclear Information System (INIS)

    Akgunduz, N.; Gimpel, R.F.; Paine, D.; Pierce, V.H.

    1997-01-01

    A one metric ton/day Vitrification Pilot Plant (VITPP) at Fernald, Ohio, simulated the vitrification of radium and radon bearing silo residues using representative non-radioactive surrogates containing high concentrations of lead, sulfates, and phosphates. The vitrification process was carried out at temperatures of 1,150 to 1,350 C. The VITPP processed glass for seven months, until a breach of the melter containment vessel suspended operations. More than 70,000 pounds of surrogate glass were produced by the VITPP. Experiences, lessons learned, and path forward will be presented

  18. Pilot plant experiments at Wairakei Power Station

    International Nuclear Information System (INIS)

    Brown, Kevin L.; Bacon, Lew G.

    2009-01-01

    In the mid-1990s, several pilot plants were constructed at Wairakei to either improve the operational and economic performance of the power station or to mitigate the environmental effects of discharges to the Waikato River. The results of the following investigations are discussed: (1) fluid flow dynamic effects on silica scaling; (2) production of silica sols of predetermined particle size to evaluate the potential for generating commercial grade silica products; (3) use of 'sulfur oxidising bacteria' for the abatement of dissolved hydrogen sulphide in cooling water; (4) removal of arsenic from separated geothermal water; (5) steam line condensate corrosion; and (6) measurement and modelling of steam scrubbing in Wairakei's long steamlines. (author)

  19. Italian experience with pilot reprocessing plants

    International Nuclear Information System (INIS)

    Cao, S.; Dworschak, H.; Rolandi, G.; Simonetta, R.

    1977-01-01

    Problems and difficulties recently experienced in the reprocessing technology of high burnup power reactor fuel elements have shown the importance of pilot plant experiments to optimize the separation processes and to test advanced equipment on a representative scale. The CNEN Eurex plant, in Saluggia (Vercelli), with a 50 kg/d thruput, in operation since '71, has completed several reprocessing campaigns on MTR type fuel elements. Two different chemical flowsheets based respectively on TBP and tertiary amines were thoroughly tested and compared: a concise comparative evaluation of the results obtained with the two schemes is given. Extensive modifications have then been introduced (namely a new headend cell equipped with a shear) to make the plant suitable to reprocess power reactor fuels. The experimental program of the plant includes a joint CNEN-AECL reprocessing experiment on CANDU (Pickering) type fuel elements to demonstrate a two cycle, amine based recovery of the plutonium. Later, a stock of high burnup fuel elements from the PWR Trino power station will be reprocessed to recover Pu and U with a Purex type flowsheet. ITREC, the second CNEN experimental reprocessing plant located at Trisaia Nuclear Center (Matera), started active operation two years ago. In the first campaign Th-U mixed oxide fuel elements irradiated in the Elk River reactor were processed. Results of this experiment are reported. ITREC special design features confer a high degree of versability to the plant allowing for substantial equipment modification under remote control conditions. For this reason the plant will be principally devoted in the near future to advanced equipment testing. Along this line high speed centrifugal contactor of a new type developed in Poland will be tested in the plant in the frame of a joint experiment between CNEN and the Polish AEC. Later on the plant program will include experimental campaign on fast reactor fuels; a detailed study on this program is in

  20. High-speed BWR power plant simulations on the special-purpose peripheral processor AD10

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1985-01-01

    A newly developed technique is described for fast, on-line simulations of normal and accidental transients in nuclear power plants. The technique is based on the utilization of the special-purpose peripheral processor AD10, which is specifically designed for high-speed systems simulations through integration of large systems of nonlinear ordinary differential equations. The Peach Bottom-II Boiling Water Reactor power plant has been simulated and results are presented. It is shown that the new technique not only advances safety analyses but also supports plant monitoring, failure diagnosis and accident mitigation, as well as the training of nuclear power plant operators. (author)

  1. Decommissioning of an uranium hexafluoride pilot plant

    International Nuclear Information System (INIS)

    Santos, Ivan; Abrao, Alcidio; Carvalho, Fatima M.S.; Ayoub, Jamil M.S.

    2009-01-01

    The Institute of Nuclear and Energetic Researches has completed fifty years of operation, belongs to the National Commission for Nuclear Energy, it is situated inside the city of Sao Paulo. The IPEN-CNEN/SP is a Brazilian reference in the nuclear fuel cycle, researches in this field began in 1970, having dominance in the cycle steps from Yellow Cake to Uranium Hexafluoride technology. The plant of Uranium Hexafluoride produced 35 metric tonnes of this gas by year, had been closed in 1992, due to domain and total transference of know-how for industrial scale, demand of new facilities for the improvement of recent researches projects. The Institute initiates decommissioning in 2002. Then, the Uranium Hexafluoride pilot plant, no doubt the most important unit of the fuel cycle installed at IPEN-CNEN/SP, beginning decommissioning and dismantlement (D and D) in 2005. Such D and D strategies, planning, assessment and execution are described, presented and evaluated in this paper. (author)

  2. DU-AGG pilot plant design study

    International Nuclear Information System (INIS)

    Lessing, P.A.; Gillman, H.

    1996-07-01

    The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule

  3. Waste Isolation Pilot Plant, Land Management Plan

    Energy Technology Data Exchange (ETDEWEB)

    1993-12-01

    To reflect the requirement of section 4 of the Wastes Isolation Pilot Plant Land Withdrawal Act (the Act) (Public Law 102-579), this land management plan has been written for the withdrawal area consistent with the Federal Land Policy and Management Act of 1976. The objective of this document, per the Act, is to describe the plan for the use of the withdrawn land until the end of the decommissioning phase. The plan identifies resource values within the withdrawal area and promotes the concept of multiple-use management. The plan also provides opportunity for participation in the land use planning process by the public and local, State, and Federal agencies. Chapter 1, Introduction, provides the reader with the purpose of this land management plan as well as an overview of the Waste Isolation Pilot Plant. Chapter 2, Affected Environment, is a brief description of the existing resources within the withdrawal area. Chapter 3, Management Objectives and Planned Actions, describes the land management objectives and actions taken to accomplish these objectives.

  4. Waste Isolation Pilot Plant, Land Management Plan

    International Nuclear Information System (INIS)

    1993-01-01

    To reflect the requirement of section 4 of the Wastes Isolation Pilot Plant Land Withdrawal Act (the Act) (Public Law 102-579), this land management plan has been written for the withdrawal area consistent with the Federal Land Policy and Management Act of 1976. The objective of this document, per the Act, is to describe the plan for the use of the withdrawn land until the end of the decommissioning phase. The plan identifies resource values within the withdrawal area and promotes the concept of multiple-use management. The plan also provides opportunity for participation in the land use planning process by the public and local, State, and Federal agencies. Chapter 1, Introduction, provides the reader with the purpose of this land management plan as well as an overview of the Waste Isolation Pilot Plant. Chapter 2, Affected Environment, is a brief description of the existing resources within the withdrawal area. Chapter 3, Management Objectives and Planned Actions, describes the land management objectives and actions taken to accomplish these objectives

  5. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde; Estudio de ruido ambiental en una planta BWR como la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  6. Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants

    International Nuclear Information System (INIS)

    Su, T.M.

    1981-05-01

    Inplant tests of safety/relief valve (SRV) discharges may be required to confirm generically established specifications for SRV loads and the maximum suppression pool temperature, and to evaluate possible effects of plant-unique parameters. These tests are required in those plants which have features that differ substantially from those previously tested. Guidelines for formulating appropriate test matrices, establishing test procedures, selecting necessary instrumentation, and reporting the test results are provided in this report. Guidelines to determine if inplant tests are required on the basis of the plant unique parameters are also included in the report

  7. Operation and maintenance support expert systems for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Miyazawa, Tatsuo; Hayakawa, Hiroyasu; Fukutomi, Shigeki

    1990-01-01

    Toshiba has been developing expert systems for operation and maintenance support in BWRs. These expert systems are designed to be integrated with conventional plant monitoring systems, and maintenance management systems to provide both comprehensive and powerful support capabilities. Some of these expert systems are already being used on a trial basis both within Toshiba and in electric power utility companies. This paper describes expert systems for the support of plant and equipment monitoring, maintenance scheduling, and maintenance work procedure planning. (author)

  8. Reliability and improvement of RODOS results for a BWR plant; Erhoehung der Zuverlaessigkeit der RODOS-Ergebnisse fuer eine SWR-Anlage

    Energy Technology Data Exchange (ETDEWEB)

    Loeffler, H.; Cester, F.; Sonnenkalb, M.; Klein-Hessling, W.; Voggenberger, T.

    2009-06-15

    Decision support systems such as RODOS aim to support the responsible authorities by providing estimates for the possible radiological consequences in case of an event in a nuclear plant. The prognosis of quantity, composition and time of occurrence of a release from the plant (''source term'') in the so-called pre-release phase is one of the foundations with high relevance for this purpose. Within previous projects source term prognosis tools have been developed and applied exemplarily for a PWR. At the end of 2005 GRS has finalized a PSA level 2 for a plant of the SWR-69 type. On this basis improved versions of the source term prognosis tools QPRO (probabilistic) and ASTRID (deterministic) have been created for a BWR and tested in an emergency exercise in a BWR. The further development of QPRO has been related in particular to the structure of the probabilistic network and the precalculated source terms. The activities for the adaptation of ASTRID focus on the creation of the dataset for the BWR coolant loop and the containment. In the emergency exercise the manageability of QPRO but also of ASTRID has been proven. Further, the first phases of the accident progression have been well identified. However, the exercise scenario developed into a very unlikely sequence with partial core melt, and the reactor building ventilation was shut off just at a critical moment. Therefore the source term prognoses deviate from the exercise scenario. Starting from these experiences with the development and application of QPRO and ASTRID recommendations are given for the further improvement of the reliability of the source term prognosis for RODOS. In general it can be stated that the development status of QPRO and ASTRID is definitely advanced compared to the presently still prevailing source term prognosis methods. Therefore it is recommended to develop plant specific versions of these codes and to apply them.

  9. TASK 3: PILOT PLANT GASIFIER TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Fusselman, Steve

    2015-11-01

    Aerojet Rocketdyne (AR) has developed an innovative gasifier concept incorporating advanced technologies in ultra-dense phase dry feed system, rapid mix injector, and advanced component cooling to significantly improve gasifier performance, life, and cost compared to commercially available state-of-the-art systems. Design, fabrication and initial testing of the pilot plant compact gasifier was completed in 2011 by a development team led by AR. Findings from this initial test program, as well as subsequent gasifier design and pilot plant testing by AR, identified a number of technical aspects to address prior to advancing into a demonstration-scale gasifier design. Key among these were an evaluation of gasifier ability to handle thermal environments with highly reactive coals; ability to handle high ash content, high ash fusion temperature coals with reliable slag discharge; and to develop an understanding of residual properties pertaining to gasification kinetics as carbon conversion approaches 99%. The gasifier did demonstrate the ability to withstand the thermal environments of highly reactive Powder River Basin coal, while achieving high carbon conversion in < 0.15 seconds residence time. Continuous operation with the high ash fusion temperature Xinyuan coal was demonstrated in long duration testing, validating suitability of outlet design as well as downstream slag discharge systems. Surface area and porosity data were obtained for the Xinyuan and Xinjing coals for carbon conversion ranging from 85% to 97%, and showed a pronounced downward trend in surface area per unit mass carbon as conversion increased. Injector faceplate measurements showed no incremental loss of material over the course of these experiments, validating the commercially traceable design approach and supportive of long injector life goals. Hybrid testing of PRB and natural gas was successfully completed over a wide range of natural gas feed content, providing test data to anchor predictions

  10. Consideration of severe accident issues for the general electric BWR standard plant a status report

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982 the U.S. NRC proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. This paper, GE's submission, discusses the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at TMI, is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  11. Facility for processing the condensates from nuclear power plants (BWR and PWR)

    International Nuclear Information System (INIS)

    Lucker, Georges.

    1975-01-01

    A plant for the processing of the condensates from boiling water or pressurized water nuclear power plants is presented. A series of couples of units for the processing of the condensates through mixed beds of ion exchange resins simultaneously ensures the filtration and demineralization of the condensates. When the resins are saturated, each mixed bed is transferred into a unit of regeneration of said resins. Each processing unit is a sphere made of a stainless material, and provided with a plurality of air and water pipes allowing the admission and evacuation of the various elements to be successively controlled [fr

  12. Pilot and pilot-commercial plants for reprocessing spent fuels of FBR type reactors

    International Nuclear Information System (INIS)

    Shaldaev, V.S.; Sokolova, I.D.

    1988-01-01

    A review of modern state of investigations on the FBR mixed oxide uranium-plutonium fuel reprocessing abroad is given. Great Britain and France occupy the leading place in this field, operating pilot plants of 5 tons a year capacity. Technology of spent fuel reprocessing and specific features of certain stages of the technological process are considered. Projects of pilot and pilot-commercial plants of Great Britain, France, Japan, USA are described. Economic problems of the FBR fuel reprocessing are touched upon

  13. BWR-plant simulator and its neural network companion with programming under mat lab environment

    International Nuclear Information System (INIS)

    Ghenniwa, Fatma Suleiman

    2008-01-01

    Stand alone nuclear power plant simulators, as well as building blocks based nuclear power simulator are available from different companies throughout the world. In this work, a review of such simulators has been explored for both types. Also a survey of the possible authoring tools for such simulators development has been performed. It is decided, in this research, to develop prototype simulator based on components building blocks. Further more, the authoring tool (Mat lab software) has been selected for programming. It has all the basic tools required for the simulator development similar to that developed by specialized companies for simulator like MMS, APROS and others. Components simulations, as well as integrated components for power plant simulation have been demonstrated. Preliminary neural network reactor model as part of a prepared neural network modules library has been used to demonstrate module order shuffling during simulation. The developed components library can be refined and extended for further development. (author)

  14. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Nonboel, E.

    1975-12-01

    A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

  15. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  16. Development of 1000kW-class MCFC pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Ooue, M.; Yasue, H. [MCFC Research Association, Mie (Japan); Takasu, K.; Tsuchitori, T.

    1996-12-31

    This pilot plant is a part of the New Sunshine Program which has proceeded by the Agency of Industrial Science and Technology of the Ministry of International Trade and Industry. MCFC Research Association is entrusted with the development of the pilot plant, and constructing it at Kawagoe site. Following items will be verified by this pilot plant operation. (a) Development of 250kW class stack and confirmation of stack performance and decay rate. (b) System verification such as basic process, control system and operation characteristics, toward commercialization. (c) To get design data for demonstration plant.

  17. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  18. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  19. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  20. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  1. Ningyo Toge uranium enrichment pilot plant comes into full

    International Nuclear Information System (INIS)

    1982-01-01

    The uranium enrichment pilot plant of the Power Reactor and Nuclear Fuel Development Corporation at Ningyo Toge went into full operation on March 26, 1982. This signifies that the front end of the nuclear fuel cycle in Japan, from uranium ore to enrichment, is only a step away from commercialization. On the same day, the pilot plant of uranium processing and conversion to UF 6 , the direct purification of uranium ore into uranium hexafluoride, began batch operation at the same works. The construction of the uranium enrichment pilot plant has been advanced in three stages: i.e. OP-1A with 1000 centrifuges, OP-1B with 3000 centrifuges and OP-2 with 3000 centrifuges. With a total of 7000 centrifuges, the pilot plant, the first enrichment plant in Japan, has now a capacity of supplying enriched uranium for six months operation of a 1,000 MW nuclear power plant. (J.P.N.)

  2. Improvements of water chemistry for moderating environmental impacts of BWR plants

    International Nuclear Information System (INIS)

    Otoha, Kei-ichi; Uetake, Naohito; Uchida, Shunsuke.

    1997-01-01

    In order to cause fewer environmental impacts, nuclear power plant systems and operational procedures with fewer radwaste sources, lower occupational exposures and smaller operational areas contaminated by radioactivity which will minimize internal exposure, have been established. Collaborative efforts have demonstrated that the combined application of major improvements in the systems and better operational procedures (e.g., (1) prevention of fuel defects, (2) application of low cobalt containing and corrosion resistant materials for the primary cooling system, (3) improvement of condensate water cleanup system and the resins applied to it, and (4) careful water chemistry control) is resulting in BWRs involved in the Japanese Improvement and Standardization Program (JISP BWRs) having occupational exposure of less than 0.5 man·Sv/yr, minimized contaminated areas in turbine buildings to only around the main steam turbines, and radwaste sources producing fewer than 500 drums/yr. Additionally, the JISP BWRs have high reliability with more than a 75% duty factor and unscheduled plant shutdown occurrences of fewer than once in 10 years. (author)

  3. Management of radioactive waste from a major core damage in a BWR power plant

    International Nuclear Information System (INIS)

    Elkert, J.; Christensen, H.; Torstenfelt, B.

    1990-01-01

    Large amounts of fission products would be released in case of a major core damage in a nuclear power reactor. In this theoretical study the core damage is caused by a loss of coolant accident followed by a complete loss of all electric power for about 30 minutes resulting in the release of 10% of the core inventory of noble gases. A second case has also been briefly studied, in which the corresponding core damage is supposed to be created merely by the complete loss of electric power during a limited time period. It appears from the study that the radioactive waste generated as a consequence of an accident of the extent can be managed in the reference reactor with only minor modifications required in the waste plant. The detailed results of the study are reactor specific, but many of the findings and recommendations are generally applicable. (author) 28 refs

  4. Pilot plant study for ethanol production

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J S [Korea Inst. of Science and Technology, Seoul (Korea, Republic of)

    1996-02-01

    Most of domestic alcohol fermentation factory adopt batch process of which productivity is lower than continuous fermentation process. They have made great effort to increase productivity by means of partial unit process automatization and process improvement with their accumulated experience but there is technical limitation in productivity of batch fermentation process. To produce and supply fuel alcohol, economic aspects must be considered first of all. Therefore, development of continuous fermentation process, of which productivity is high, is prerequisite to produce and use fuel alcohol but only a few foreign company possess continuous fermentation technic and use it in practical industrial scale fermentation. We constructed pilot plant (5 Stage CSTR 1 kl 99.5 v/v% ethanol/Day scale) to study some aspects stated below and our ultimate aims are production of industrial scale fuel alcohol and construction of the plant by ourselves. Some study concerned with energy saving separation and contamination control technic were entrusted to KAIST, A-ju university and KIST respectively. (author) 67 refs., 100 figs., 58 tabs.

  5. Draft environmental assessment: Ocean Thermal Energy Conversion (OTEC) Pilot Plants

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, S.M.; Sands, M.D.; Donat, J.R.; Jepsen, P.; Smookler, M.; Villa, J.F.

    1981-02-01

    This Environmental Assessment (EA) has been prepared, in accordance with the National Environmental Policy Act of 1969, for the deployment and operation of a commercial 40-Megawatt (MW) Ocean Thermal Energy Conversion (OTEC) Pilot Plant (hereafter called the Pilot Plant). A description of the proposed action is presented, and a generic environment typical of the candidate Pilot Plant siting regions is described. An assessment of the potential environmental impacts associated with the proposed action is given, and the risk of credible accidents and mitigating measures to reduce these risks are considered. The Federal and State plans and policies the proposed action will encompass are described. Alternatives to the proposed action are presented. Appendix A presents the navigation and environmental information contained in the US Coast Pilot for each of the candidate sites; Appendix B provides a brief description of the methods and calculations used in the EA. It is concluded that environmental disturbances associated with Pilot Plant activities could potentially cause significant environmental impacts; however, the magnitude of these potential impacts cannot presently be assessed, due to insufficient engineering and environmental information. A site- and design-specific OTEC Pilot Plant Environmental Impact Statement (EIS) is required to resolve the potentially significant environmental effects associated with Pilot Plant deployment and operation. (WHK)

  6. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Rodenas, J.; Abarca, A.; Gallardo, S.

    2011-01-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool.

  7. Waste Isolation Pilot Plant No-migration variance petition

    International Nuclear Information System (INIS)

    1990-03-01

    This report describes various aspects of the Waste Isolation Pilot Plant (WIPP) including design data, waste characterization, dissolution features, ground water hydrology, natural resources, monitoring, general geology, and the gas generation/test program

  8. Waste Isolation Pilot Plant Salt Decontamination Testing

    Energy Technology Data Exchange (ETDEWEB)

    Rick Demmer; Stephen Reese

    2014-09-01

    On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

  9. Waste isolation pilot plant disposal room model

    Energy Technology Data Exchange (ETDEWEB)

    Butcher, B.M.

    1997-08-01

    This paper describes development of the conceptual and mathematical models for the part of the Waste Isolation Pilot Plant (WIPP) repository performance assessment that is concerned with what happens to the waste over long times after the repository is decommissioned. These models, collectively referred to as the {open_quotes}Disposal Room Model,{close_quotes} describe the repository closure process during which deformation of the surrounding salt consolidates the waste. First, the relationship of repository closure to demonstration of compliance with the Environmental Protection Agency (EPA) standard (40 CFR 191 Appendix C) and how sensitive performance results are to it are examined. Next, a detailed description is provided of the elements of the disposal region, and properties selected for the salt, waste, and other potential disposal features such as backfill. Included in the discussion is an explanation of how the various models were developed over time. Other aspects of closure analysis, such as the waste flow model and method of analysis, are also described. Finally, the closure predictions used in the final performance assessment analysis for the WIPP Compliance Certification Application are summarized.

  10. Waste Isolation Pilot Plant Strategic Plan

    International Nuclear Information System (INIS)

    1993-03-01

    The purpose of the Waste Isolation Pilot Plant (WIPP) Strategic Plan is to provide decision makers, project participants, and the public with a high-level overview of the objectives, issues, and strategiesthat impact a decision on the suitability of WIPP as a permanent, safe disposal facility for transuranic (TRU) waste that has resulted from defense activities. This document is a component of an integrated planning process and is a key management tool that is coordinated and consistent with the Secretary's Disposal Decision Plan and the Environmental Restoration and Waste Management (EM) Five-Year Plan. This documentsupports other US Department of Energy (DOE) planning efforts, including the TRU Waste Program. The WIPP Strategic Plan addresses the WIPP Program Test Phase, Disposal Decision, Disposal Phase, and Decommissioning Phase (decontamination and decommissioning). It describes the actions and activities that the DOE will conduct to ensure that WIPP will comply with applicable, relevant, and appropriate requirements of the US Environmental Protection Agency (EPA), State of New Mexico, and other applicable federal and state regulations. It also includes the key assumptions under which the strategy was developed. A comprehensive discussion of the multitude of activities involved in the WIPP Program cannot be adequately presented in this document. The specific details of these activities are presented in other, more detailed WIPP planningdocuments

  11. Waste Isolation Pilot Plant borehole data

    International Nuclear Information System (INIS)

    1995-04-01

    Data pertaining to all the surface boreholes used at the WIPP site for site characterization hydrological testing and resource evaluation exist in numerous source documents. This project was initiated to develop a comprehensive data base that would include the data on all WIPP related surface boreholes from the Atomic Energy Commission, Waste Isolation Pilot Plant Energy Research and Development Administration, Department of Energy, and Hydrologic Test Borehole Programs. The data compiled from each borehole includes: operator, permit number, location, total depth, type of well, driller, drilling record, casing record, plugging schedule, and stratigraphic summary. There are six groups of boreholes contained in this data base, they are as follows: Commercially Drilled Potash Boreholes, Energy Department Wells, Geologic Exploration Boreholes, Hydrologic Test Boreholes, Potash Boreholes, and Subsurface Exploration Boreholes. There were numerous references which contained borehole data. In some cases the data found in one document was inconsistent with data in another document. In order to ensure consistency and accuracy in the data base, the same references were used for as many of the boreholes as possible. For example, all elevations and locations were taken from Compilation and Comparison of Test-Hole Location Surveys in the Vicinity of the WIPP Site. SAND 88-1065, Table 3-5. There are some sections where a data field is left blank. In this case, the information was either not applicable or was unavailable

  12. Waste isolation pilot plant disposal room model

    International Nuclear Information System (INIS)

    Butcher, B.M.

    1997-08-01

    This paper describes development of the conceptual and mathematical models for the part of the Waste Isolation Pilot Plant (WIPP) repository performance assessment that is concerned with what happens to the waste over long times after the repository is decommissioned. These models, collectively referred to as the open-quotes Disposal Room Model,close quotes describe the repository closure process during which deformation of the surrounding salt consolidates the waste. First, the relationship of repository closure to demonstration of compliance with the Environmental Protection Agency (EPA) standard (40 CFR 191 Appendix C) and how sensitive performance results are to it are examined. Next, a detailed description is provided of the elements of the disposal region, and properties selected for the salt, waste, and other potential disposal features such as backfill. Included in the discussion is an explanation of how the various models were developed over time. Other aspects of closure analysis, such as the waste flow model and method of analysis, are also described. Finally, the closure predictions used in the final performance assessment analysis for the WIPP Compliance Certification Application are summarized

  13. Waste Isolation Pilot Plant Environmental Monitoring Plan

    International Nuclear Information System (INIS)

    2008-01-01

    U.S. Department of Energy (DOE) Order 450.1, Environmental Protection Program, requires each DOE site to conduct environmental monitoring. Environmental monitoring at the Waste Isolation Pilot Plant (WIPP) is conducted in order to: (a) Verify and support compliance with applicable federal, state, and local environmental laws, regulations, permits, and orders; (b) Establish baselines and characterize trends in the physical, chemical, and biological condition of effluent and environmental media; (c) Identify potential environmental problems and evaluate the need for remedial actions or measures to mitigate the problems; (d) Detect, characterize, and report unplanned releases; (e) Evaluate the effectiveness of effluent treatment and control, and pollution abatement programs; and (f) Determine compliance with commitments made in environmental impact statements, environmental assessments, safety analysis reports, or other official DOE documents. This Environmental Monitoring Plan (EMP) explains the rationale and design criteria for the environmental monitoring program, extent and frequency of monitoring and measurements, procedures for laboratory analyses, quality assurance (QA) requirements, program implementation procedures, and direction for the preparation and disposition of reports. Changes to the environmental monitoring program may be necessary to allow the use of advanced technology and new data collection techniques. This EMP will document changes in the environmental monitoring program. Guidance for preparation of EMPs is contained in DOE/EH-0173T, Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance.

  14. Waste Isolation Pilot Plant safety analysis report

    International Nuclear Information System (INIS)

    1997-03-01

    The United States Department of Energy (DOE) was authorized by Public Law 96-164 to provide a research and development facility for demonstrating the safe permanent disposal of transuranic (TRU) wastes from national defense activities and programs of the United States exempted from regulations by the US Nuclear Regulatory Commission (NRC). The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico near Carlsbad, was constructed to determine the efficacy of an underground repository for disposal of TRU wastes. In accordance with the 1981 and 1990 Records of Decision (ROD), the development of the WIPP was to proceed with a phased approach. Development of the WIPP began with a siting phase, during which several sites were evaluated and the present site selected based on extensive geotechnical research, supplemented by testing. The site and preliminary design validation phase (SPDV) followed the siting phase, during which two shafts were constructed, an underground testing area was excavated, and various geologic, hydrologic, and other geotechnical features were investigated. The construction phase followed the SPDV phase during which surface structures for receiving waste were built and underground excavations were completed for waste emplacement

  15. Waste Isolation Pilot Plant (WIPP) startup plan

    International Nuclear Information System (INIS)

    1988-03-01

    To allow the Waste Isolation Pilot Plant (WIPP) to transition from a Major System Acquisition to an operating demonstration facility, the Acquisition Executive and the Energy System Acquisition Advisory Board (ESAAB) must concur in the facility's readiness to receive waste. This action, designated in DOE Order 4700.1 as Key Decision Four, concludes with the Chairman of the ESAAB issuing a Record of Decision. Since the meeting leading to the Record of Decision is scheduled for August 1988, plans must be made to ensure all activities contributing to that decision are completed in a clear and well-coordinated process. To support that effort, this Start-Up Plan was prepared to identify and track key events necessary to verify WIPP's readiness to receive waste; this provides a management/scheduling/tracking tool for the DOE WIPP Project Office (WPO) and a tracking mechanism for the DOE Albuquerque Operations Office (DOE-AL) and for DOE Headquarters (DOE-HQ); and describe the process to ensure readiness is documented by providing relevant data and reports to the cognizant decision makers. The methods by which these two purposes are achieved are discussed in further detail in the remainder of this plan

  16. Waste Isolation Pilot Plant borehole data

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    Data pertaining to all the surface boreholes used at the WIPP site for site characterization hydrological testing and resource evaluation exist in numerous source documents. This project was initiated to develop a comprehensive data base that would include the data on all WIPP related surface boreholes from the Atomic Energy Commission, Waste Isolation Pilot Plant Energy Research and Development Administration, Department of Energy, and Hydrologic Test Borehole Programs. The data compiled from each borehole includes: operator, permit number, location, total depth, type of well, driller, drilling record, casing record, plugging schedule, and stratigraphic summary. There are six groups of boreholes contained in this data base, they are as follows: Commercially Drilled Potash Boreholes, Energy Department Wells, Geologic Exploration Boreholes, Hydrologic Test Boreholes, Potash Boreholes, and Subsurface Exploration Boreholes. There were numerous references which contained borehole data. In some cases the data found in one document was inconsistent with data in another document. In order to ensure consistency and accuracy in the data base, the same references were used for as many of the boreholes as possible. For example, all elevations and locations were taken from Compilation and Comparison of Test-Hole Location Surveys in the Vicinity of the WIPP Site. SAND 88-1065, Table 3-5. There are some sections where a data field is left blank. In this case, the information was either not applicable or was unavailable.

  17. Particle collection by a pilot plant venturi scrubber downstream from a pilot plant electrostatic precipitator

    Science.gov (United States)

    Sparks, L. E.; Ramsey, G. H.; Daniel, B. E.

    The results of pilot plant experiments of particulate collection by a venturi scrubber downstream from an electrostatic precipitator (ESP) are presented. The data, which cover a range of scrubber operating conditions and ESP efficiencies, show that particle collection by the venturi scrubber is not affected by the upstream ESP; i.e., for a given scrubber pressure drop, particle collection efficiency as a function of particle diameter is the same for both ESP on and ESP off. The experimental results are in excellent agreement with theoretical predictions. Order of magnitude cost estimates indicate that particle collection by ESP scrubber systems may be economically attractive when scrubbers must be used for SO x control.

  18. Pilot plant for flue gas treatment - continuous operation tests

    International Nuclear Information System (INIS)

    Chmielewski, A.G.; Tyminski, B.; Iller, E.; Zimek, Z.; Licki, J.; Radzio, B.

    1995-01-01

    Tests of continuous operation have been performed on pilot plant at EPS Kaweczyn in the wide range of SO 2 concentration (500-3000 ppm). The bag filter has been applied for aerosol separation. The high efficiencies of SO 2 and NO x removal, approximately 90% were obtained and influenced by such process parameters as: dose, gas temperature and ammonia stoichiometry. The main apparatus of the pilot plant (e.g. both accelerators) have proved their reliability in hard industrial conditions. (Author)

  19. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  20. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  1. Waste water pilot plant research, development, and demonstration permit application

    International Nuclear Information System (INIS)

    1993-03-01

    This permit application has been prepared to obtain a research, development, and demonstration permit to perform pilot-scale treatability testing on the 242-A Evaporator process condensate waste water effluent stream. It provides the management framework, and controls all the testing conducted in the waste water pilot plant using dangerous waste. It also provides a waste acceptance envelope (upper limits for selected constituents) and details the safety and environmental protection requirements for waste water pilot plant testing. This permit application describes the overall approach to testing and the various components or requirements that are common to all tests. This permit application has been prepared at a sufficient level of detail to establish permit conditions for all waste water pilot plant tests to be conducted

  2. Waste Isolation Pilot Plant Environmental Monitoring Plan

    International Nuclear Information System (INIS)

    2004-01-01

    U.S. Department of Energy (DOE) Order 450.1, Environmental Protection Program, requires each DOE site to conduct environmental monitoring. Environmental monitoring at the Waste Isolation Pilot Plant (WIPP) is conducted in order to: (a) Verify and support compliance with applicable federal, state, and local environmental laws, regulations, permits, and orders; (b) Establish baselines and characterize trends in the physical, chemical, and biological condition of effluent and environmental media; (c) Identify potential environmental problems and evaluate the need for remedial actions or measures to mitigate the problem; (d) Detect, characterize, and report unplanned releases; (e) Evaluate the effectiveness of effluent treatment and control, and pollution abatement programs; and (f) Determine compliance with commitments made in environmental impact statements, environmental assessments, safety analysis reports, or other official DOE documents. This Environmental Monitoring Plan (EMP) has been written to contain the rationale and design criteria for the monitoring program, extent and frequency of monitoring and measurements, procedures for laboratory analyses, quality assurance (QA) requirements, program implementation procedures, and direction for the preparation and disposition of reports. Changes to the environmental monitoring program may be necessary to allow the use of advanced technology and new data collection techniques. This EMP will document any proposed changes in the environmental monitoring program. Guidance for preparation of Environmental Monitoring Plans is contained in DOE/EH-0173T, Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance. The plan will be effective when it is approved by the appropriate Head of Field Organization or their designee. The plan discusses major environmental monitoring and hydrology activities at the WIPP and describes the programs established to ensure that WIPP operations do not

  3. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  4. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  5. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  6. Encapsulation pilot plant of radioactive wastes in thermosetting resins

    International Nuclear Information System (INIS)

    1982-01-01

    The thermosetting resins (polyesters, epoxides) are used to encapsulate the low and intermediate - level radioactive wastes. The testing program concerning the drums produced by the pilot plant of the Chooz nuclear power plant is described. The installation operating is examined while thinking of the industrial application. The production costs are then evaluated

  7. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  9. FGD Franchising Pilot Project of Thermal Power Plants

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    According to the national policy on enhancing environmental protection,the five major power generation companies are required to carry out flue gas desulphurization(FGD) franchising pilot project in thermal power plants.This paper introduces the development of this pilot project,including the foundation,purpose,objects,demands and procedures.It also discusses some main problems encountered during implementation,involving the understanding,legislation,financing,taxation,pricing and management of franchise.At...

  10. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  11. Correlating activity incorporation with properties of oxide films formed on material samples exposed to BWR and PWR coolants in Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P. [VTT Industrial Systems, Espoo (Finland); Buddas, T.; Halin, M.; Kvarnstroem, R.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant, Loviisa (Finland); Helin, M.; Muttilainen, E.; Reinvall, A. [Teollisuuden Voima Oy, Olkiluoto (Finland)

    2002-07-01

    The extent of activity incorporation on primary circuit surfaces in nuclear power plants is connected to the chemical composition of the coolant, to the corrosion behaviour of the material surfaces and to the structure and properties of oxide films formed on circuit surfaces due to corrosion. Possible changes in operational conditions may induce changes in the structure of the oxide films and thus in the rate of activity incorporation. To predict these changes, experimental correlations between water chemistry, oxide films and activity incorporation, as well as mechanistic understanding of the related phenomena need to be established. In order to do this, flow-through cells with material samples and facilities for high-temperature water chemistry monitoring have been installed at Olkiluoto unit 1 (BWR) and Loviisa unit 1 (PWR) in spring 2000. The cells are being used for two major purposes: To observe the changes in the structure and activity levels of oxide films formed on material samples exposed to the primary coolant. Correlating these observations with the abundant chemical and radiochemical data on coolant composition, dose rates etc. collected routinely by the plant, as well as with high-temperature water chemistry monitoring data such as the corrosion potentials of relevant material samples, the redox potential and the high-temperature conductivity of the primary coolant. We describe in this paper the scope of the work, give examples of the observations made and summarize the results on oxide films that have been obtained during one full fuel cycle at both plants. (authors)

  12. Photovoltaic pilot plant of Vulcano (Italy)

    Energy Technology Data Exchange (ETDEWEB)

    Menga, P.; Previ, A.

    1988-12-31

    The rated peak power of the Vulcano plant is about 80 kW. The main feature of the plant lies in its dual operating mode: either stand-alone or grid-connected. The plant, commissioned half-way through 1984, has always worked uninterruptedly and satisfactorily and, for this reason, it is expected to keep on operating in the coming years. Beginning from 1987, it has been considered worth carrying out research on a number of specific points not so far dealt with in sufficient detail in literature, as follows: experimental research on the deterioration, over a period of time, of the characteristics of the modules of the plant; theoretical and experimental research on the plant`s electrochemical storage system. The research on the second of the above subjects has a number of aims: to improve methods of checking the state of charge of batteries in newly-designed photovoltaic plants; to improve the ways in which the Vulcano battery storage system is operated; and to monitor and make a careful diagnosis of the condition of the plant`s battery.

  13. Electroosmotically enhanced sludge dewatering-pilot-plant study

    CSIR Research Space (South Africa)

    Smollen, M

    1994-01-01

    Full Text Available role in determining the ease or difficulty of phase separation. It seems that the inefficiency of dewatering applied to gelatinous and fine-particle sludges can be overcome by mechanical dewatering enhanced by electroosmosis. A prototype pilot-plant...

  14. Dissolution studies with pilot plant and actual INTEC calcines

    International Nuclear Information System (INIS)

    Herbst, R.S.; Garn, T.G.

    1999-01-01

    The dissolution of Idaho Nuclear Technology and Engineering Center (INTEC) pilot plant calcines was examined to determine solubility of calcine matrix components in acidic media. Two representatives pilot plant calcine types were studied: Zirconia calcine and Zirconia/Sodium calcine. Dissolution of these calcines was evaluated using lower initial concentrations of nitric acid than used in previous tests to decrease the [H+] concentration in the final solutions. Lower [H+] concentrations contribute to more favorable TRUEX/SREX solvent extraction flowsheet performance. Dissolution and analytical results were also obtained for radioactive calcines produced using high sodium feeds blended with non-radioactive Al(NO 3 ) 3 solutions to dilute the sodium concentration and prevent bed agglomeration during the calcination process. Dissolution tests indicated >95 wt.% of the initial calcine mass can be dissolved using the baseline dissolution procedure, with the exception that higher initial nitric acid concentrations are required. The higher initial acid concentration is required for stoichiometric dissolution of the oxides, primarily aluminum oxide. Statistically designed experiments using pilot plant calcine were performed to determine the effect of mixing rate on dissolution efficiency. Mixing rate was determined to provide minimal effects on wt.% dissolution. The acid/calcine ratio and temperature were the predominate variables affecting the wt.% dissolution, a result consistent with previous studies using other similar types of pilot plant calcines

  15. Rock mechanics activities at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Francke, C.; Saeb, S.

    1996-01-01

    The application of rock mechanics at nuclear waste repositories is a true multidisciplinary effort. A description and historical summary of the Waste Isolation Pilot Plant (WIPP) is presented. Rock mechanics programs at the WIPP are outlined, and the current rock mechanics modeling philosophy of the Westinghouse Waste Isolation Division is discussed

  16. Waste Isolation Pilot Plant Technical Assessment Team Report

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2015-03-17

    This report provides the results of the Waste Isolation Pilot Plant (WIPP) technical assessment led by the Savannah River National Laboratory and conducted by a team of experts in pertinent disciplines from SRNL and Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), Pacific Northwest National Laboratory (PNNL), and Sandia National Laboratories (SNL).

  17. Analysis of Waste Isolation Pilot Plant Samples: Integrated Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Britt, Phillip F [ORNL

    2015-03-01

    Analysis of Waste Isolation Pilot Plant Samples: Integrated Summary Report. Summaries of conclusions, analytical processes, and analytical results. Analysis of samples taken from the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico in support of the WIPP Technical Assessment Team (TAT) activities to determine to the extent feasible the mechanisms and chemical reactions that may have resulted in the breach of at least one waste drum and release of waste material in WIPP Panel 7 Room 7 on February 14, 2014. This report integrates and summarizes the results contained in three separate reports, described below, and draws conclusions based on those results. Chemical and Radiochemical Analyses of WIPP Samples R-15 C5 SWB and R16 C-4 Lip; PNNL-24003, Pacific Northwest National Laboratory, December 2014 Analysis of Waste Isolation Pilot Plant (WIPP) Underground and MgO Samples by the Savannah River National Laboratory (SRNL); SRNL-STI-2014-00617; Savannah River National Laboratory, December 2014 Report for WIPP UG Sample #3, R15C5 (9/3/14); LLNL-TR-667015; Lawrence Livermore National Laboratory, January 2015 This report is also contained in the Waste Isolation Pilot Plant Technical Assessment Team Report; SRNL-RP-2015-01198; Savannah River National Laboratory, March 17, 2015, as Appendix C: Analysis Integrated Summary Report.

  18. Pilot plant for the radioactive decontamination of spent oils

    International Nuclear Information System (INIS)

    Flores E, R.M.; Ortiz O, H.V.; Cisneros L, L.; Lopez G, R.

    2002-01-01

    In this work the operation parameters obtained in the laboratory of oil storage are presented, as well as the operations which shape the pilot plant, the design criteria and the basic design of the core equipment of the developed process. Finally, the comparative results obtained the decontamination process of oil are given as well as laboratory scale. (Author)

  19. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  20. Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2017-01-01

    Full Text Available Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR. Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

  1. Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

    Science.gov (United States)

    Tanaka, Ken-ichi; Ueno, Jun

    2017-09-01

    Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

  2. Identification of Radioactive Pilot-Plant test requirements

    Energy Technology Data Exchange (ETDEWEB)

    Powell, W.J.; Riebling, E.F.

    1995-05-09

    Radioactive Pilot-Plant testing needs and alternatives are evaluated for enhanced Sludge Washing and High and Low-Level Vitrification efforts. Also investigated was instrument and equipment testing needs associated with the vitrification and retrieval process. The scope of this document is to record the existing March 1994 letter report for future use. A structured Kepner-Trego{trademark} decision analysis process was used to assist analysis of the testing needs. This analysis provided various combinations of laboratory and radioactive (hot) and cold pilot testing options associated with the above need areas. Recommendations for testing requirements were made.

  3. Identification of Radioactive Pilot-Plant test requirements

    International Nuclear Information System (INIS)

    Powell, W.J.; Riebling, E.F.

    1995-01-01

    Radioactive Pilot-Plant testing needs and alternatives are evaluated for enhanced Sludge Washing and High and Low-Level Vitrification efforts. Also investigated was instrument and equipment testing needs associated with the vitrification and retrieval process. The scope of this document is to record the existing March 1994 letter report for future use. A structured Kepner-Trego trademark decision analysis process was used to assist analysis of the testing needs. This analysis provided various combinations of laboratory and radioactive (hot) and cold pilot testing options associated with the above need areas. Recommendations for testing requirements were made

  4. The pilot plant for electron beam food processing

    International Nuclear Information System (INIS)

    Migdal, W.; Kosmal, W.; Malec-Czechowska, K.; Maciszewski, W.

    1992-01-01

    In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in INCT. The pilot plant has been constructed inside an old fort what decreases significantly the cost of the investment. The pilot plants is equipped with a small research accelerator Pilot (10 MeV, 1 kW) and an industrial unit Elektronika (10 MeV, 10 kW). This allows both laboratory and full technological scale testing of the elaborated process to be conducted. The industrial unit is being equipped with e-/X conversion target, for high density products irradiation. On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permissions for permanent treatment of spices, garlic, onions and temporary permissions for mushrooms, and potatoes. Dosimetric methods have been elaborated for the routine use at the plant. In the INCT laboratory methods for the control of e-/X treated food have been established. (author). 9 refs, 5 figs, 1 tab

  5. The pilot plant for electron beam food processing

    Science.gov (United States)

    Migdal, W.; Walis, L.; Chmielewski, A. G.

    1993-07-01

    In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in INCT. The pilot plant has been constructed inside an old fort what decreases significantly the cost of the investment. The pilot plant is equipped with a small research accelerator Pilot (10 MeV, 1 kW) and an industrial unit Elektronika (10 MeV, 10 kW). This allows both laboratory and full technological scale testing of the elaborated process to be conducted. The industrial unit is being equipped with e-/X conversion target, for high density products irradiation. On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permissions for permanent treatment of spices, garlic, onions and temporary permissions for mushrooms, and potatoes. Dosimetric methods have been elaborated for the routine use at the plant. In the INCT laboratory methods for the control of e-/X treated food have been established.

  6. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  7. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  8. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  9. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  10. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  11. Pilot plant of continuous ion-exchange in fluidized bed

    International Nuclear Information System (INIS)

    Botella, T.; Otero de Becerra, J.; Gasos, P.

    1985-01-01

    Research and development on continuous ion-exchange processes has been a major item in hydrometallurgy. This new technology has been under development during the last 15 years in the leading countries at uranium hydrometallurgy. The fluidized bed multi-stage column technique is proven to be the most attractive one, and since 1977 several commercial plants have begun production, some of them with a throughput of 500 cubic meters of pregnant liquour per hour. J.E.N. undertook the study of this new technology for uranium recovery in the early 70's. In 1979 a pilot plant had been installed, based on previous laboratory and smaller pilot plant experience. The plant was designed following JEN's own technology and has been operating successfully at a flow rate of near 0.5 cubic meters per hour. The test runs and the main processing, engineering and operation features are described. At present a demonstation plant is under design, and this installation will provide the necessary know-how for the construction and operation of a commercial scale plant. (author)

  12. MBR pilot plant for textile wastewater treatment and reuse.

    Science.gov (United States)

    Lubello, C; Caffaz, S; Mangini, L; Santianni, D; Caretti, C

    2007-01-01

    An experimental study was carried out in order to evaluate the possibility of upgrading the conventional activated sludge WWTP of Seano (Prato, Italy) which treats municipal and textile wastewaters, by using membrane bioreactor (MBR) technology. The MBR pilot plant, set up within Seano WWTP, was fed with mixed municipal-industrial wastewaters during the first experimental period and with pure industrial wastewaters during the second. Performances and operation of the MBR were evaluated in terms of permeate characteristics and variability (COD, colour, surfactants, total N and P) and other operational parameters (sludge growth and observed yield). According to the experimental results the MBR permeate quality was always superior to the Seano WWTP one and it was suitable for industrial reuse in the textile district of the Prato area. Respirometric tests provided a modified IWA ASM1 model which fits very well the experimental data and can be used for the design and the monitoring of a full-scale MBR pilot plant.

  13. Process control of an HTGR fuel reprocessing cold pilot plant

    International Nuclear Information System (INIS)

    Rode, J.S.

    1976-10-01

    Development of engineering-scale systems for a large-scale HTGR fuel reprocessing demonstration facility is currently underway in a cold pilot plant. These systems include two fluidized-bed burners, which remove the graphite (carbon) matrix from the crushed HTGR fuel by high temperature (900 0 C) oxidation. The burners are controlled by a digital process controller with an all analog input/output interface which has been in use since March, 1976. The advantages of such a control system to a pilot plant operation can be summarized as follows: (1) Control loop functions and configurations can be changed easily; (2) control constants, alarm limits, output limits, and scaling constants can be changed easily; (3) calculation of data and/or interface with a computerized information retrieval system during operation are available; (4) diagnosis of process control problems is facilitated; and (5) control panel/room space is saved

  14. A shaft seal system for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Hansen, F.D.; Ahrens, E.H.; Dennis, A.W.; Hurtado, L.D.; Knowles, M.K.; Tillerson, J.R.; Thompson, T.W.; Galbraith, D.

    1996-01-01

    As part of the demonstration of compliance with federal regulations, a shaft seal system has been designed for the Waste Isolation Pilot Plant. The system completely fills the 650 m shafts with components consisting of the common engineering materials, each of which possesses low permeability, longevity, and can be constructed using available technology. Design investigations couple rock mechanics and fluid flow analysis and tests of these materials within the natural geological setting, and demonstrate the effectiveness of the design

  15. Pilot-plant development of a Rover waste calcination flowsheet

    International Nuclear Information System (INIS)

    Birrer, S.A.

    1978-04-01

    Results of eight runs, six using the 10-cm dia and two using the 30-cm dia pilot-plant calciners, in which simulated first-cycle Rover waste was calcined, are described. Results of the tests showed that a feed blend consisting of one volume simulated first-cycle Rover waste and one or two volumes simulated first-cycle zirconium waste could not be successfully calcined. 5 figs., 8 tables

  16. Continuous precipitation of uranium peroxide in process pilot plant

    International Nuclear Information System (INIS)

    Quinelato, A.L.

    1990-01-01

    An experimental study on uranium peroxide precipitation has been carried out with the objective to evaluate the influence of the main process parameters with a technological approach. The uraniferous solution used was obtained from the hydrometallurgical processing of an ore from Itataia - CE. Studies were developed in two distinct experimental stages. In the first stage, the precipitation was investigated by means of laboratory batch tests and, in the second stage, by means of continuous operation in a process pilot plant. (author)

  17. Solar Pilot Plant project review No. 9, May 4--5, 1977. CDRL item 10

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-01-01

    Drawings and illustrations for the project review are presented. These are included for the 10 MW(e) solar pilot plant, the collector subsystem, the receiver subsystem, the electrical power generation system and balance of plant, plant controls and transient analysis, availability and safety, pilot and commercial plant designs, and summary and recommendations. (MHR)

  18. Technical Integration of SMART Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Park, P. H.; Noh, P. C. (and others)

    2006-12-15

    Preliminary experimental tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual heat removal (PRHR) systems. The integrated SMART desalination plant consists of Multi Effect Distillation Process combined with Thermal-Vapor Compressor(MED-TVC) and coupled with the extracted steam from turbine through the steam transformer. Steam transformer produces the main pressure steam and supplies to the MED-TVC unit. MED-TVC was selected as a desalination process coupled with SMART, since the thermal vapor compression is very effective where the steam is available at high temperature and pressure conditions than required in the evaporator. The standard design of the SMART desalination plant is under development as a part of the SMART project. This report describes design concept of these systems and their requirements.

  19. First geothermal pilot power plant in Hungary

    Directory of Open Access Journals (Sweden)

    Tóth Anikó

    2007-01-01

    Full Text Available The Hungarian petroleum industry has always participated in the utilization of favourable geothermal conditions in the country. Most of the Hungarian geothermal wells were drilled by the MOL Ltd. as CH prospect holes. Accordingly, the field of geothermics belonged to the petroleum engineering, although marginally. It was therefore a surprise to hear of the decision of MOL Ltd. to build a geothermal power plant of about 2-5 MW. The tender was published in 2004.The site selected for the geothermal project is near the western border of an Hungarian oilfield, close to the Slovenian border. The location of the planned geothermal power plant was chosen after an analysis of suitable wells owned by the MOL Rt. The decision was made on the bases of different reservoir data. The existence of a reservoir of the necessary size, temperature, permeability, productivity and the water chemistry data was proved. The wells provide an enough information to understand the character of the reservoir and will be the production wells used by the planned power plant.The depth of the wells is about 2930 - 3200 m. The Triassic formation is reached at around 2851 m. The production and the reinjection wells are planned. The primary objective of the evaluation is to further learn the nature of the geothermal system. First a one-day discharge test is carried out. If this short-term test is successful, a six-months long-term discharge test will follow. The first period of the test is a transient phenomenon. Within the well test, the wellhead pressure, the flow rate, the outflowing water temperature, the dynamic fluid level, and the chemical components will be measured. The heat transfer around the bore-hole is influenced by the flow rate and the time. For the right appreciation of the measured data, it is very important to analyse the heat transfer processes around the bore-hole. The obtained data from the experiments must be also fitted into the framework of a mathematical

  20. Pilot plant for hydrogasification of coal with nuclear heat

    International Nuclear Information System (INIS)

    Falkenhain, G.; Velling, G.

    1976-01-01

    In the framework of a research and development programme sponsored by the Ministry of Research and Technology of the Federal Republic of Germany, two process variants for hydrogasification of coal by means of nuclear heat have been developed by the Rheinische Braunkohlenwerke AG, Cologne. For testing these process variants a semi-technical pilot plant for gasification of coal under pressure in a fluidized bed was constructed. The pilot plant, in which the gasification of lignite and hard coal is planned, is designed for a throughput of 100kg carbon per hour corresponding to 400kg raw lignite per hour or 150kg hard coal per hour. The plant should provide data on the influence of the most essential process parameters (pressure, temperature, residence time of gas and coal, type and pre-treatment of feed coal) on the performance of gasification and raw gas composition. Different plant components will also be tested. Since the pilot plant will permit testing of both process variants of hydrogasification, it was designed in such a way that it is possible to vary a great number of process parameters. Thus, for instance, the pressure can be chosen in a range up to 100 bar and pure hydrogen or mixtures of hydrogen, carbon monoxide and steam can be applied as gasification agents. The gasifier is an internally insulated fluidized bed reactor with an inner diameter of 200mm and a height of about 8m, to which an internally insulated cyclone for separation of the entrained fines is attached. The raw gas is then cooled down by direct water scrubbing. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  2. Radiological effects of a nuclear power plant on a river system, as demonstrated by the Gundremmingen BWR on the Danube

    International Nuclear Information System (INIS)

    Herrmann, H.; Ruf, M.; Huebel, K.; Luensmann, W.

    1975-01-01

    The Gundremmingen Nuclear Power Plant (boiling water reactor, 237 MW(e)), on the Upper Danube, has been in operation since 1967. Radiological data have been collected throughout its period of operation. The behaviour of the radioactive waste products in the ecological system of the Upper Danube (water, sediments, suspended solids, water-plants, fish) has been analysed in connection with environmental contamination, the uptake capacity of the Danube, and the possible pathways to man. As a result of the investigations, it seems possible to build further nuclear power plants on the Danube, if their rates of release of radioactivity are similar to those at Gundremmingen. (author)

  3. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  4. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  5. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

    International Nuclear Information System (INIS)

    2005-10-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

  6. Experimental results: Pilot plant calcine dissolution and liquid feed stability

    International Nuclear Information System (INIS)

    Herbst, R.S.; Fryer, D.S.; Brewer, K.N.; Johnson, C.K.; Todd, T.A.

    1995-02-01

    The dissolution of simulated Idaho Chemical Processing Plant pilot plant calcines, containing none of the radioactive actinides, lanthanides or fission products, was examined to evaluate the solubility of calcine matrix materials in acidic media. This study was a necessary precursor to dissolution and optimization experiments with actual radionuclide-containing calcines. The importance of temperature, nitric acid concentration, ratio of acid volume to calcine mass, and time on the amount, as a weight percentage of calcine dissolved, was evaluated. These parameters were studied for several representative pilot plant calcine types: (1) Run No. 74 Zirconia calcine; (2) Run No. 17 Zirconia/Sodium calcine; (3) Run No. 64 Zirconia/Sodium calcine; (3) Run No. 1027 Alumina calcine; and (4) Run No. 20 Alumina/Zirconia/Sodium calcine. Statistically designed experiments with the different pilot plant calcines indicated the effect of the studied process variables on the amount of calcine dissolved decreases in the order: Acid/Calcine Ratio > Temperature > HNO 3 Concentration > Dissolution Time. The following conditions are suitable to achieve greater than 90 wt. % dissolution of most Zr, Al, or Na blend calcines: (1) Maximum nitric acid concentration of 5M; (2) Minimum acid/calcine ratio of 10 mL acid/1 gram calcine; (3) Minimum dissolution temperature of 90 degrees C; and (4) Minimum dissolution time of 30 minutes. The formation of calcium sulphate (CaSO 4 ) precipitates was observed in certain dissolved calcine solutions during the dissolution experiments. Consequently, a study was initiated to evaluate if and under what conditions the resulting dissolved calcine solutions would be unstable with regards to precipitate formation. The results indicate that precipitate formation in the calcine solutions prepared under the above proposed dissolution conditions are not anticipated

  7. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

    International Nuclear Information System (INIS)

    2005-10-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and ware out of components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of heavy water moderated reactors (HWRs), boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

  8. Methodology to classify accident sequences of an Individual Plant Examination according to the severe releases for BWR type reactors

    International Nuclear Information System (INIS)

    Sandoval V, S.

    2001-01-01

    The Light Water Reactor (LWR) operation regulations require to every operating plant to perform of an Individual Plant Examination study (Ipe). One of the main purposes of an Ipe is t o gain a more quantitative understanding of the overall probabilities of core damage and fission product releases . Probabilistic Safety Analysis (PSA) methodologies and Severe Accident Analysis are used to perform Ipe studies. PSA methodologies are used to identify and analyse the set of event sequences that might originate the fission product release from a nuclear power plant; these methodologies are combinatorial in nature and generate thousands of sequences. Among other uses within an Ipe, severe accident simulations are used to determine the characteristics of the fission product release for the identified sequences and in this way, the releases can be understood and characterized. A vast amount of resources is required to simulate and analyse every Ipe sequence. This effort is unnecessary if similar sequences are grouped. The grouping scheme must achieve an efficient trade off between problem reduction and accuracy. The methodology presented in this work enables an accurate characterization and analysis of the Ipe fission product releases by using a reduced problem. The methodology encourages the use of specific plant simulations. (Author)

  9. List of key words with classification for a standard safety report for nuclear power plants with PWR or BWR

    International Nuclear Information System (INIS)

    1976-01-01

    Under the efforts of improving the licensing procedure for nuclear power plants, the Federal Minister of the Interior set up a task group of experts of the manufacturers and operators of nuclear power plants, the assessors (Technische Ueberwachungsvereine, TUeVs), the Institute for Reactor Safety of the TUeVs, the licensing authorities of the Laender, and the Federal Ministry of the Interior which worked out a list of key words for writing the safety report for nuclear power plants with PWRs and BWRs. This list of key words is published herewith in order to encourage its application when writing or assessing safety reports for nuclear power plants and in order to present the opportunity to make proposals for improvement to a group as large as possible. At a later date, it is intended to incorporate the list of key words as soon as sufficient experience from the practical application will justify this, it is intended to incorporate the list of key words in a general administrative regulation. (orig.) [de

  10. Technical Proposal Salton Sea Geothermal Power Pilot Plant Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1975-03-28

    The proposed Salton Sea Geothermal Power Pilot Plant Program comprises two phases. The objective of Phase 1 is to develop the technology for power generation from high-temperature, high-salinity geothermal brines existing in the Salton Sea known geothermal resources area. Phase 1 work will result in the following: (a) Completion of a preliminary design and cost estimate for a pilot geothermal brine utilization facility. (b) Design and construction of an Area Resource Test Facility (ARTF) in which developmental geothermal utilization concepts can be tested and evaluated. Program efforts will be divided into four sub-programs; Power Generation, Mineral Extraction, Reservoir Production, and the Area Resources Test Facility. The Power Generation Subprogram will include testing of scale and corrosion control methods, and critical power cycle components; power cycle selection based on an optimization of technical, environmental and economic analyses of candidate cycles; preliminary design of a pilot geothermal-electric generating station to be constructed in Phase 2 of this program. The Mineral Extraction Subprogram will involve the following: selection of an optimum mineral recovery process; recommendation of a brine clean-up process for well injection enhancement; engineering, construction and operation of mineral recovery and brine clean-up facilities; analysis of facility operating results from environmental, economical and technical point-of-view; preliminary design of mineral recovery and brine clean-up facilities of sufficient size to match the planned pilot power plant. The Reservoir Production Subprogram will include monitoring the operation and maintenance of brine production, handling and injection systems which were built with private funding in phase 0, and monitoring of the brine characteristics and potential subsidence effects during well production and injection. Based on the above, recommendations and specifications will be prepared for production and

  11. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  12. Pressurized fluidized bed combustion combined cycle power plant with coal gasification: Second generation pilot plant

    International Nuclear Information System (INIS)

    Farina, G.L.; Bressan, L.

    1991-01-01

    This paper presents the technical and economical background of a research and development program of a novel power generation scheme, which is based on coal gasification, pressurized fluid bed combustion and combined cycles. The participants in this program are: Foster Wheeler (project leader), Westinghouse, IGT and the USA Dept. of Energy. The paper describes the characteristics of the plant, the research program in course of implementation, the components of the pilot plant and the first results obtained

  13. Waste Isolation Pilot Plant CY 2000 Site Environmental Report

    International Nuclear Information System (INIS)

    Westinghouse TRU Solutions, LLC; Environmental Science and Research Foundation, Inc.

    2001-01-01

    The U.S. Department of Energy's (DOE) Carlsbad Field Office and Westinghouse TRU Solutions, LLC (WTS) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environmental, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 2000 Site Environmental Report summarizes environmental data from calendar year (CY) 2000 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH-0173T), and the Waste Isolation Pilot Plant Environmental Protect ion Implementation Plan (DOE/WIPP 96-2199). The above orders and guidance documents require that DOE facilities submit an Annual Site Environmental Report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health. The purpose of this report is to provide a comprehensive description of operational environmental monitoring activities, to provide an abstract of environmental activities conducted to characterize site environmental management performance to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit at WIPP during CY 2000. The format of this report follows guidance offered in a June 1, 2001 memo from DOE's Office of Policy and Guidance with the subject ''Guidance for the preparation of Department of Energy (DOE) Annual Site Environmental Reports (ASERs) for Calendar Year 2000.'' WIPP received its first shipment of waste on March 26, 1999. In 2000, no evidence was found of any adverse

  14. Waste Isolation Pilot Plant 2001 Site Environmental Report

    International Nuclear Information System (INIS)

    Westinghouse TRU Solutions, Inc.

    2002-01-01

    The United States (U.S.) Department of Energy's (DOE) Carlsbad Field Office (CBFO) and Westinghouse TRU Solutions LLC (WTS) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environmental, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 2001 Site Environmental Report summarizes environmental data from calendar year (CY) 2001 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH- 0173T), and the Waste Isolation Pilot Plant Environmental Protection Implementation Plan (DOE/WIPP 96-2199). The above Orders and guidance documents require that DOE facilities submit an annual site environmental report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health; and the New Mexico Environment Department (NMED). The purpose of this report is to provide a comprehensive description of operational environmental monitoring activities, to provide an abstract of environmental activities conducted to characterize site environmental management performance to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit at WIPP during CY 2001. WIPP received its first shipment of waste on March 26, 1999. In 2001, no evidence was found of any adverse effects from WIPP on the surrounding environment

  15. Waste Isolation Pilot Plant CY 2000 Site Environmental Report

    Energy Technology Data Exchange (ETDEWEB)

    Westinghouse TRU Solutions, LLC; Environmental Science and Research Foundation, Inc.

    2001-12-31

    The U.S. Department of Energy's (DOE) Carlsbad Field Office and Westinghouse TRU Solutions, LLC (WTS) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environmental, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 2000 Site Environmental Report summarizes environmental data from calendar year (CY) 2000 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH-0173T), and the Waste Isolation Pilot Plant Environmental Protect ion Implementation Plan (DOE/WIPP 96-2199). The above orders and guidance documents require that DOE facilities submit an Annual Site Environmental Report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health. The purpose of this report is to provide a comprehensive description of operational environmental monitoring activities, to provide an abstract of environmental activities conducted to characterize site environmental management performance to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit at WIPP during CY 2000. The format of this report follows guidance offered in a June 1, 2001 memo from DOE's Office of Policy and Guidance with the subject ''Guidance for the preparation of Department of Energy (DOE) Annual Site Environmental Reports (ASERs) for Calendar Year 2000.'' WIPP received its first shipment of waste on March 26, 1999. In 2000, no

  16. Waste Isolation Pilot Plant 1999 Site Environmental Report

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Roy B.; Adams, Amy; Martin, Don; Morris, Randall C.; Reynolds, Timothy D.; Warren, Ronald W.

    2000-09-30

    The U.S. Department of Energy's (DOE)Carlsbad Area Office and the Westinghouse Waste Isolation Division (WID) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environmental, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 1999 Site Environmental Report summarizes environmental data from calendar year 1999 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH- 0173T), and the Waste Isolation Pilot Plant Environmental Protection Implementation Plan (DOE/WIPP 96-2199). The above orders and guidance documents require that DOE facilities submit an Annual Site Environmental Report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health. The purpose of this report is to provide a comprehensive description of operational environmental monitoring activities, to provide an abstract of environmental activities conducted to characterize site environmental management performance to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit at WIPP during calendar year 1999. WIPP received its first shipment of waste on March 26, 1999. In 1999, no evidence was found of any adverse effects from WIPP on the surrounding environment. Radionuclide concentrations in the environment surrounding WIPP were not statistically higher in 1999 than in 1998.

  17. Waste Isolation Pilot Plant 2001 Site Environmental Report

    Energy Technology Data Exchange (ETDEWEB)

    Westinghouse TRU Solutions, Inc.

    2002-09-20

    The United States (U.S.) Department of Energy's (DOE) Carlsbad Field Office (CBFO) and Westinghouse TRU Solutions LLC (WTS) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environmental, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 2001 Site Environmental Report summarizes environmental data from calendar year (CY) 2001 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH- 0173T), and the Waste Isolation Pilot Plant Environmental Protection Implementation Plan (DOE/WIPP 96-2199). The above Orders and guidance documents require that DOE facilities submit an annual site environmental report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health; and the New Mexico Environment Department (NMED). The purpose of this report is to provide a comprehensive description of operational environmental monitoring activities, to provide an abstract of environmental activities conducted to characterize site environmental management performance to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit at WIPP during CY 2001. WIPP received its first shipment of waste on March 26, 1999. In 2001, no evidence was found of any adverse effects from WIPP on the surrounding environment.

  18. Water Treatment Pilot Plant Design Manual: Low Flow Conventional/Direct Filtration Water Treatment Plant for Drinking Water Treatment Studies

    Science.gov (United States)

    This manual highlights the project constraints and concerns, and includes detailed design calculations and system schematics. The plant is based on engineering design principles and practices, previous pilot plant design experiences, and professional experiences and may serve as ...

  19. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  20. Geotechnical Perspectives on the Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    Francke, Chris T.; Hansen, Frank D.; Knowles, M. Kathyn; Patchet, Stanley J.; Rempe, Norbert T.

    1999-01-01

    The Waste Isolation Pilot Plant (WIPP) is the first nuclear waste repository certified by the United States Environmental Protection Agency. Success in regulatory compliance resulted from an excellent natural setting for such a repository, a facility with multiple, redundant safety systems, and from a rigorous, transparent scientific and technical evaluation. The WIPP story, which has evolved over the past 25 years, has generated a library of publications and analyses. Details of the multifaceted program are contained in the cited references. Selected geotechnical highlights prove the eminent suitability of the WIPP to serve its congressionally mandated purpose

  1. Hydrologic studies for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Davies, P.B.

    1994-01-01

    The objective of this paper is to provide a general overview of hydrologic conditions at the Waste Isolation Pilot Plant (WIPP) by describing several key hydrologic studies that have been carried out as part of the site characterization program over the last 20 years. The paper is composed of three parts: background information about general objectives of the WIPP project; information about the geologic and hydrologic setting of the facility; and information about three aspects of the hydrologic system that are important to understanding the long-term performance of the WIPP facility. For additional detailed information, the reader is referred to the references cited in the text

  2. The Waste Isolation Pilot Plant status and related socioeconomic impacts

    International Nuclear Information System (INIS)

    Little, C.C.; Adcock, L.D.; Hohmann, G.L.

    1984-01-01

    The Waste Isolation Pilot Plant (WIPP) has been ''authorized as a defense activity of the Department of Energy...for the express purpose of providing a research and development facility to demonstrate the safe disposal of radioactive wastes resulting from the defense activities and programs of the United States...'' (PL 96-164). As reported in previous conferences, WIPP continues ahead of schedule and below budget with full facility construction well underway. To date, based on recent review, the socioeconomic impacts have been negligible and steps have been taken to ensure that they remain that way throughout operations

  3. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    Energy Technology Data Exchange (ETDEWEB)

    Washington Regulatory and Environmental Services (WRES)

    2004-10-25

    This Biennial Environmental Compliance Report (BECR) documents environmental regulatory compliance at the Waste Isolation Pilot Plant (WIPP), a facility designed and authorized for the safe disposal of transuranic (TRU) radioactive waste, for the reporting period of April 1, 2002, to March 31, 2004. As required by the WIPP Land Withdrawal Act (LWA) (Public Law [Pub. L.] 102-579, as amended by Pub. L. 104-201), the BECR documents U.S. Department of Energy (DOE) compliance with applicable environmental protection laws and regulations implemented by agencies of the federal government and the state of New Mexico.

  4. Operations Program Plan for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1990-09-01

    This document, Revision 4 of the Operations Program Plan, has been developed as the seven-year master plan for operating of the Waste Isolation Pilot Plant (WIPP). Subjects covered include public and technical communications; regulatory and environmental programs; startup engineering; radiation handling, surface operations, and underground operations; waste certification and waste handling; transportation development; geotechnical engineering; experimental operations; engineering program; general maintenance; security program; safety, radiation, and regulatory assurance; quality assurance program; training program; administration activities; management systems program; and decommissioning. 243 refs., 19 figs., 25 tabs. (SM)

  5. Waste Isolation Pilot Plant Annual Site Environmental Report for 2012

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-09-01

    The purpose of the Waste Isolation Pilot Plant (WIPP) Annual Site Environmental Report for 2012 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1B, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to: Characterize site environmental management performance; Summarize environmental occurrences and responses reported during the calendar year; Confirm compliance with environmental standards and requirements; Highlight significant environmental accomplishments, including progress toward the DOE Environmental Sustainability Goals made through implementation of the WIPP Environmental Management System (EMS).

  6. The waste isolation pilot plant regulatory compliance program

    International Nuclear Information System (INIS)

    Mewhinney, J.A.; Kehrman, R.F.

    1996-01-01

    The passage of the WIPP Land Withdrawal Act of 1992 (LWA) marked a turning point for the Waste Isolation Pilot Plant (WIPP) program. It established a Congressional mandate to open the WIPP in as short a time as possible, thereby initiating the process of addressing this nation's transuranic (TRU) waste problem. The DOE responded to the LWA by shifting the priority at the WIPP from scientific investigations to regulatory compliance and the completion of prerequisites for the initiation of operations. Regulatory compliance activities have taken four main focuses: (1) preparing regulatory submittals; (2) aggressive schedules; (3) regulator interface; and (4) public interactions

  7. High Temperature Calcination - MACT Upgrade Equipment Pilot Plant Test

    Energy Technology Data Exchange (ETDEWEB)

    Richard D. Boardman; B. H. O& #39; Brien; N. R. Soelberg; S. O. Bates; R. A. Wood; C. St. Michel

    2004-02-01

    About one million gallons of acidic, hazardous, and radioactive sodium-bearing waste are stored in stainless steel tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), which is a major operating facility of the Idaho National Engineering and Environmental Laboratory. Calcination at high-temperature conditions (600 C, with alumina nitrate and calcium nitrate chemical addition to the feed) is one of four options currently being considered by the Department of Energy for treatment of the remaining tank wastes. If calcination is selected for future processing of the sodium-bearing waste, it will be necessary to install new off-gas control equipment in the New Waste Calcining Facility (NWCF) to comply with the Maximum Achievable Control Technology (MACT) standards for hazardous waste combustors and incinerators. This will require, as a minimum, installing a carbon bed to reduce mercury emissions from their current level of up to 7,500 to <45 {micro}g/dscm, and a staged combustor to reduce unburned kerosene fuel in the off-gas discharge to <100 ppm CO and <10 ppm hydrocarbons. The staged combustor will also reduce NOx concentrations of about 35,000 ppm by 90-95%. A pilot-plant calcination test was completed in a newly constructed 15-cm diameter calciner vessel. The pilot-plant facility was equipped with a prototype MACT off-gas control system, including a highly efficient cyclone separator and off-gas quench/venturi scrubber for particulate removal, a staged combustor for unburned hydrocarbon and NOx destruction, and a packed activated carbon bed for mercury removal and residual chloride capture. Pilot-plant testing was performed during a 50-hour system operability test January 14-16, followed by a 100-hour high-temperature calcination pilot-plant calcination run January 19-23. Two flowsheet blends were tested: a 50-hour test with an aluminum-to-alkali metal molar ratio (AAR) of 2.25, and a 50-hour test with an AAR of 1.75. Results of the testing

  8. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    International Nuclear Information System (INIS)

    2004-01-01

    This Biennial Environmental Compliance Report (BECR) documents environmental regulatory compliance at the Waste Isolation Pilot Plant (WIPP), a facility designed and authorized for the safe disposal of transuranic (TRU) radioactive waste, for the reporting period of April 1, 2002, to March 31, 2004. As required by the WIPP Land Withdrawal Act (LWA) (Public Law [Pub. L.] 102-579, as amended by Pub. L. 104-201), the BECR documents U.S. Department of Energy (DOE) compliance with applicable environmental protection laws and regulations implemented by agencies of the federal government and the state of New Mexico.

  9. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  10. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  11. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  12. Modeling temperature variations in a pilot plant thermophilic anaerobic digester.

    Science.gov (United States)

    Valle-Guadarrama, Salvador; Espinosa-Solares, Teodoro; López-Cruz, Irineo L; Domaschko, Max

    2011-05-01

    A model that predicts temperature changes in a pilot plant thermophilic anaerobic digester was developed based on fundamental thermodynamic laws. The methodology utilized two simulation strategies. In the first, model equations were solved through a searching routine based on a minimal square optimization criterion, from which the overall heat transfer coefficient values, for both biodigester and heat exchanger, were determined. In the second, the simulation was performed with variable values of these overall coefficients. The prediction with both strategies allowed reproducing experimental data within 5% of the temperature span permitted in the equipment by the system control, which validated the model. The temperature variation was affected by the heterogeneity of the feeding and extraction processes, by the heterogeneity of the digestate recirculation through the heating system and by the lack of a perfect mixing inside the biodigester tank. The use of variable overall heat transfer coefficients improved the temperature change prediction and reduced the effect of a non-ideal performance of the pilot plant modeled.

  13. Waste Isolation Pilot Plant Site Environmental Report Calendar Year 2002

    International Nuclear Information System (INIS)

    Washington Regulatory and Environmental Services

    2003-01-01

    The United States (U.S.) Department of Energy (DOE) Carlsbad Field Office (CBFO) and Washington TRU Solutions LLC (WTS) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environment, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 2002 Site Environmental Report summarizes environmental data from calendar year 2002 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, and Guidance for the Preparation of DOE Annual Site Environmental Reports (ASERs) for Calendar Year 2002 (DOE Memorandum EH-41: Natoli:6-1336, April 4, 2003). These Orders and the guidance document require that DOE facilities submit an annual site environmental report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health; and the New Mexico Environment Department (NMED)

  14. Waste Isolation Pilot Plant Site Environmental Report Calendar Year 2002

    Energy Technology Data Exchange (ETDEWEB)

    Washington Regulatory and Environmental Services

    2003-09-17

    The United States (U.S.) Department of Energy (DOE) Carlsbad Field Office (CBFO) and Washington TRU Solutions LLC (WTS) are dedicated to maintaining high quality management of Waste Isolation Pilot Plant (WIPP) environmental resources. DOE Order 5400.1, General Environmental Protection Program, and DOE Order 231.1, Environment, Safety, and Health Reporting, require that the environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This Waste Isolation Pilot Plant 2002 Site Environmental Report summarizes environmental data from calendar year 2002 that characterize environmental management performance and demonstrate compliance with federal and state regulations. This report was prepared in accordance with DOE Order 5400.1, DOE Order 231.1, and Guidance for the Preparation of DOE Annual Site Environmental Reports (ASERs) for Calendar Year 2002 (DOE Memorandum EH-41: Natoli:6-1336, April 4, 2003). These Orders and the guidance document require that DOE facilities submit an annual site environmental report to DOE Headquarters, Office of the Assistant Secretary for Environment, Safety, and Health; and the New Mexico Environment Department (NMED).

  15. Arizona Public Service - Alternative Fuel (Hydrogen) Pilot Plant Design Report

    Energy Technology Data Exchange (ETDEWEB)

    James E. Francfort

    2003-12-01

    Hydrogen has promise to be the fuel of the future. Its use as a chemical reagent and as a rocket propellant has grown to over eight million metric tons per year in the United States. Although use of hydrogen is abundant, it has not been used extensively as a transportation fuel. To assess the viability of hydrogen as a transportation fuel and the viability of producing hydrogen using off-peak electric energy, Pinnacle West Capital Corporation (PNW) and its electric utility subsidiary, Arizona Public Service (APS) designed, constructed, and operates a hydrogen and compressed natural gas fueling station—the APS Alternative Fuel Pilot Plant. This report summarizes the design of the APS Alternative Fuel Pilot Plant and presents lessons learned from its design and construction. Electric Transportation Applications prepared this report under contract to the U.S. Department of Energy’s Advanced Vehicle Testing Activity. The Idaho National Engineering and Environmental Laboratory manages these activities for the Advanced Vehicle Testing Activity.

  16. The Stored Waste Examination Pilot Plant program at the INEL

    International Nuclear Information System (INIS)

    McKinley, K.B.; Anderson, B.C.; Clements, T.L.; Hinckley, J.P.; Mayberry, J.L.; Smith, T.H.

    1983-01-01

    Since 1970, defense transuranic waste has been placed into 20-year retrievable storage at the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory (INEL). A major objective of the U.S. Department of Energy (DOE) Nuclear Waste Management Program is to remove all retrievably stored transuranic waste from the INEL. The January 1981 DOE Record of Decision on the Waste Isolation Pilot Plant (WIPP) stated, ''The WIPP facility will dispose of defense transuranic waste stored retrievably at the Idaho National Engineering Laboratory.'' After retrieval and before shipment, processing may be necessary to prepare the waste for acceptance, handling, and enhanced long-term isolation in the WIPP. However, some of the waste is certifiable to the WIPP waste acceptance criteria without container opening or waste processing. To minimize costs, the Stored Waste Examination Pilot Plant (SWEPP) is being developed to certify INEL stored transuranic waste without container opening or waste processing. The SWEPP certification concept is based on records assessment, nondestructive examination techniques, assay techniques, health physics examinations, and limited opening of containers at another facility for quality control

  17. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  18. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  19. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  20. Guidebook on design, construction and operation of pilot plants for uranium ore processing

    International Nuclear Information System (INIS)

    1990-01-01

    The design, construction and operation of a pilot plant are often important stages in the development of a project for the production of uranium concentrates. Since building and operating a pilot plant is very costly and may not always be required, it is important that such a plant be built only after several prerequisites have been met. The main purpose of this guidebook is to discuss the objectives of a pilot plant and its proper role in the overall project. Given the wide range of conditions under which a pilot plant may be designed and operated, it is not possible to provide specific details. Instead, this book discusses the rationale for a pilot plant and provides guidelines with suggested solutions for a variety of problems that may be encountered. This guidebook is part of a series of Technical Reports on uranium ore processing being prepared by the IAEA's Division of Nuclear Fuel Cycle and Waste Management. 42 refs, 7 figs, 3 tabs

  1. Prototype plant for nuclear process heat (PNP) - operation of the pilot plant for hydrogasification of coal

    International Nuclear Information System (INIS)

    Bruengel, N.; Dehms, G.; Fiedler, P.; Gerigk, H.P.; Ruddeck, W.; Schrader, L.; Schumacher, H.J.

    1988-04-01

    The Rheinische Braunkohlenwerke AG developed the process of hydrogasification of coal in a fluidized bed for generation of SNG. On basis of test results obtained in a semi-technical pilot plant of a through-put of 250 kg/h dried coal a large pilot plant was erected processing 10 t/h dried brown coal. This plant was on stream for about 14700 h, of which about 7800 h were with gasifier operation; during this time about 38000 t of dried brown coal of the Rhenish district were processed containing 4 to 25% of ash. At pressures of 60 to 120 bar and temperatures of 800 to 935 0 C carbon conversion rates up to 81 percent and methane amounts of 5000 m 3 (STP)/h were reached. The decisive parameter for methane generation was the hydrogen/coal-ratio. Even at high moisture contents, usually diminishing the methane yield from the coal essentially, by high hydrogen/coal-ratios high methane yields could be obtained. The gasifier itself caused no troubles during the total time operation. Difficulties with the original design of the residual char cooler could be overcome by change-over from water injection to liquid carbon dioxide. The design of the heat recovery system proved well. Alltogether so the size increasement of the gasifier from the semi-technical to the large pilot plant as well as the harmonization of gas generation and gas refining was proved. (orig.) With 20 refs., 20 tabs., 81 figs [de

  2. Waste Isolation Pilot Plant Safety Analysis Report. Volume 5

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection; Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating controls and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  3. Waste Isolation Pilot Plant Safety Analysis Report. Volume 1

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection: Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating control and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  4. Waste Isolation Pilot Plant Safety Analysis Report. Volume 4

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection; Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating controls and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  5. Waste Isolation Pilot Plant Safety Analysis Report. Volume 2

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection; Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating controls and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  6. Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

    International Nuclear Information System (INIS)

    Toft, P.; Borring, J.; Adolph, E.

    1988-01-01

    A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

  7. The significance of the pilot conditioning plant (PKA) for spent fuel management

    International Nuclear Information System (INIS)

    Willax, H.O.

    1996-01-01

    The pilot conditioning plant (PKA) is intended as a multi-purpose facility and thus may serve various purposes involved in the conditioning or disposal of spent fuel elements or radwaste. Its design as a pilot plant permits development and trial of various methods and processes for fuel element conditioning, as well as for radwaste conditioning. (orig./DG) [de

  8. General Atomic HTGR fuel reprocessing pilot plant: results of initial sequential equipment operation

    International Nuclear Information System (INIS)

    1978-09-01

    In September 1977, the processing of 20 large high-temperature gas-cooled reactor (LHTGR) fuel elements was completed sequentially through the head-end cold pilot plant equipment. This report gives a brief description of the equipment and summarizes the results of the sequential operation of the pilot plant. 32 figures, 15 tables

  9. Tetrafluoride uranium pilot plant in operation at IEA, using the moving bed process

    International Nuclear Information System (INIS)

    Franca Junior, J.M.

    1975-01-01

    A UF 4 pilot plant, in operation at IEA, using the moving bed process is reported. UO 3 obtained from the thermal decomposition of ADU is used as a starting material in this pilot plant. The type of equipment and the process are both described. Ammonia gas (NH 3 ) was used in the reduction operation and anhydrous hydrofluoric acid (HF) in the hydrofluorination step

  10. Sharing five years of pilot plant experience on aromatics extraction with ionic liquids

    NARCIS (Netherlands)

    Onink, S.A.F.; Hansmeier, A.R.; Meindersma, G.W.; Haan, de A.B.

    2011-01-01

    Since 2004 pilot plant trials have been conducted with various contactors and different ionic liquids for petrochemical model feeds as well as real refinery feeds. Our pilot plant contains several columns (rotating disc contactor, Kuhni, pulsed disc and donut column) with a height of 6 m and 5 cm

  11. Long term pilot plant experience on aromatics extraction with ionic liquids

    NARCIS (Netherlands)

    Meindersma, W.G.W.; Onink, F.S.A.F.; Hansmeier, A.R.; Haan, de A.B.

    2012-01-01

    Since 2004, we have been conducting pilot plant trials with various contactors and different ionic liquids for petrochemical model feeds as well as real refinery feeds. Our pilot plant contains a Rotating Disc Contactor with a height of 6 m and a diameter of 60 mm. Up to 100 kg of ionic liquid and

  12. Water chemistry control practices and data of the European BWR fleet

    International Nuclear Information System (INIS)

    Stellwag, B.; Laendner, A.; Weiss, S.; Huettner, F.

    2010-01-01

    Nineteen BWR plants are in operation in Europe, nine built by ASEA Atom, six by Siemens KWU and four by General Electric. This paper gives an overview of water chemistry operation practices and parameters of the European BWR plants. General design characteristics of the plants are described. Chemistry control strategies and underlying water chemistry guidelines are summarized. Chemistry data are presented and discussed with regard to plant design characteristics. The paper is based on a contract of the European BWR Forum with AREVA on a chemistry sourcebook for member plants. The survey of chemistry data was conducted for the years 2002 to 2008. (author)

  13. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  14. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  15. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  16. Common Cause Failure Analysis of Control Rods and Drives in the Swedish and Finnish BWR Plants. Operating Experiences in 1983 - 2003

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, Tuomas [Avaplan Oy, Espoo (Finland)

    2006-11-15

    substantial time difference and/or spatial distance within the core. The exploration of CCF cases showed that the most prevalent factor in the CCF mechanisms was the time coupling by sequencing maintenance into refueling outages. Essential contributing factors were design changes, deviations or errors in maintenance or new types of replacement parts, accompanied by unexpected influences. An evident positive trend could be observed both for single failures and CCFs . Impact Vectors were used to expresses the conditional failure probability for the various multiplicity in CCF events, linking event analysis to the estimation of CCF model parameters. A reference application was made for the Forsmark 1 and 2 plant. The Common Load Model was used as parametric CCF model, which proved to be a practicable approach. This method provides a consistent handling of failure combinations and workable extension to evaluate localized dependence between adjacent control rod and drives. Also international experience and reference information were surveyed. The developed methods and collected data are utilized in the ongoing PSA updates for the Swedish BWRs and Olkiluoto 1 and 2. Review - within the project a detailed and project extern review has been performed, covering also the older CCF events. This do now guarantee that the CCF data for the control rods and drives in Swedish and Finnish BWR:s during the observation period 1983 - 2003, now can be judged as quality assured. The scope of this project was limited to collection, analysis and classification of CCF data, and reference application using the industry average of pooled data. It has not been the scope of this project to perform more comprehensive probabilistic studies on e.g., positive learning trends, impact of plant specific design details or different amount of failing control rods at different operational conditions in the reactor vessel and with different safety and support systems in operation. It has either been the scope to

  17. Bentonite as a waste isolation pilot plant shaft sealing material

    International Nuclear Information System (INIS)

    Daemen, J.; Ran, Chongwei

    1996-12-01

    Current designs of the shaft sealing system for the Waste Isolation Pilot Plant (WIPP) propose using bentonite as a primary sealing component. The shaft sealing designs anticipate that compacted bentonite sealing components can perform through the 10,000-year regulatory period and beyond. To evaluate the acceptability of bentonite as a sealing material for the WIPP, this report identifies references that deal with the properties and characteristics of bentonite that may affect its behavior in the WIPP environment. This report reviews published studies that discuss using bentonite as sealing material for nuclear waste disposal, environmental restoration, toxic and chemical waste disposal, landfill liners, and applications in the petroleum industry. This report identifies the physical and chemical properties, stability and seal construction technologies of bentonite seals in shafts, especially in a saline brine environment. This report focuses on permeability, swelling pressure, strength, stiffness, longevity, and densification properties of bentonites

  18. Bentonite as a waste isolation pilot plant shaft sealing material

    Energy Technology Data Exchange (ETDEWEB)

    Daemen, J.; Ran, Chongwei [Univ. of Nevada, Reno, NV (United States)

    1996-12-01

    Current designs of the shaft sealing system for the Waste Isolation Pilot Plant (WIPP) propose using bentonite as a primary sealing component. The shaft sealing designs anticipate that compacted bentonite sealing components can perform through the 10,000-year regulatory period and beyond. To evaluate the acceptability of bentonite as a sealing material for the WIPP, this report identifies references that deal with the properties and characteristics of bentonite that may affect its behavior in the WIPP environment. This report reviews published studies that discuss using bentonite as sealing material for nuclear waste disposal, environmental restoration, toxic and chemical waste disposal, landfill liners, and applications in the petroleum industry. This report identifies the physical and chemical properties, stability and seal construction technologies of bentonite seals in shafts, especially in a saline brine environment. This report focuses on permeability, swelling pressure, strength, stiffness, longevity, and densification properties of bentonites.

  19. Waste Isolation Pilot Plant Groundwater Protection Management Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2002-09-24

    U.S. Department of Energy (DOE) Order 5400.1, General Environmental Protection Program, requires each DOE site to prepare a Groundwater Protection Management Program Plan. This document fulfills the requirement for the Waste Isolation Pilot Plant (WIPP). This document was prepared by the Hydrology Section of the Westinghouse TRU Solutions LLC (WTS) Environmental Compliance Department, and it is the responsibility of this group to review the plan annually and update it every three years. This document is not, nor is it intended to be, an implementing document that sets forth specific details on carrying out field projects or operational policy. Rather, it is intended to give the reader insight to the groundwater protection philosophy at WIPP.

  20. Final environmental impact statement. Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1980-10-01

    This volume contains the appendices for the Final Environmental Impact Statement for the Waste Isolation Pilot Plant (WIPP). Alternative geologic environs are considered. Salt, crystalline rock, argillaceous rock, and tuff are discussed. Studies on alternate geologic regions for the siting of WIPP are reviewed. President Carter's message to Congress on the management of radioactive wastes and the findings and recommendations of the interagency review group on nuclear waste management are included. Selection criteria for the WIPP site including geologic, hydrologic, tectonic, physicochemical compatability, and socio-economic factors are presented. A description of the waste types and the waste processing procedures are given. Methods used to calculate radiation doses from radionuclide releases during operation are presented. A complete description of the Los Medanos site, including archaeological and historic aspects is included. Environmental monitoring programs and long-term safety analysis program are described

  1. Waste Isolation Pilot Plant Annual Site Environmental Report for 2016

    Energy Technology Data Exchange (ETDEWEB)

    Ward, Anderson [Carlsbad Field Office (CBFO), NM (United States); Basabilvazo, George T. [Carlsbad Field Office (CBFO), NM (United States)

    2017-09-01

    The purpose of the Waste Isolation Pilot Plant (WIPP) Annual Site Environmental Report for 2016 (ASER) is to provide the information required by U.S. Department of Energy (DOE) Order 231.1B, Environment, Safety, and Health Reporting. The DOE Carlsbad Field Office (CBFO) and the management and operating contractor (MOC) maintain and preserve the environmental resources at the WIPP facility. DOE Order 231.1B; DOE Order 436.1, Departmental Sustainability; and DOE Order 458.1, Radiation Protection of the Public and the Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and workers, and preservation of the environment. This report was prepared in accordance with DOE Order 231.1B, which requires DOE facilities to submit an ASER to the DOE Headquarters Chief Health, Safety, and Security Officer.

  2. Waste Isolation Pilot Plant Groundwater Protection Management Program Plan

    International Nuclear Information System (INIS)

    1993-01-01

    The DOE has mandated in DOE Order 5400.1 that its operations will be conducted in an environmentally safe manner. The Waste Isolation Pilot Plant (WIPP) will comply with DOE Order 5400.1 and will conduct its operations in a manner that ensures the safety of the environment and the public. This document outlines how the WIPP will protect and preserve groundwater within and surrounding the WIPP facility. Groundwater protection is just one aspect of the WIPP environmental protection effort. The WIPP groundwater surveillance program is designed to determine statistically if any changes are occurring in groundwater characteristics within and surrounding the WIPP facility. If a change is noted, the cause will be determined and appropriate corrective action initiated

  3. Annual stability evaluation of Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1993-06-01

    A stability evaluation of the underground workings of the Waste Isolation Pilot Plant (WIPP) was completed by the US Bureau of Mines' WIPP evaluation committee. This work included a critical evaluation of the processes employed at WIPP to ensure stability, an extensive review of available deformation measurements, a 3-day site visit, and interviews with the Department of Energy (DOE) and Westinghouse staff. General ground control processes are in place at WIPP to minimize the likelihood that major stability problems will go undetected. To increase confidence in both short- and long-term stability throughout the site (underground openings and shafts), ground stability monitoring systems, mine layout design, support systems and data analyses must be continuously improved. Such processes appear to be in place at WIPP and are discussed in this paper

  4. Cold flow model study of an oxyfuel combustion pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Guio-Perez, D.C.; Tondl, G.; Hoeltl, W.; Proell, T.; Hofbauer, H. [Vienna University of Technology, Institute of Chemical Engineering, Vienna (Austria)

    2011-12-15

    The fluid-dynamic behavior of a circulating fluidized bed pilot plant for oxyfuel combustion was studied in a cold flow model, down-scaled using Glicksman's criteria. Pressures along the unit and the global circulation rate were used for characterization. The analysis of five operating parameters and their influence on the system was carried out; namely, total solids inventory and the air velocity of primary, secondary, loop seal and support fluidizations. The cold flow model study shows that the reactor design allows stable operation at a wide range of fluidization rates, with results that agree well with previous observations described in the literature. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  5. Revised concept for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Dennis, A.W.; Milloy, J.A.; Scully, L.W.; Shefelbine, H.C.; Stinebaugh, R.E.; Wowak, W.E.

    1978-07-01

    The quantities of remotely handled wastes that must be handled at the Waste Isolation Pilot Plant have been reduced from 250 x 10 3 ft 3 /y to 10 x 10 3 ft 3 /y; the capital cost of the facility will be reduced from 534 to 428 million dollars. Changes in the facility design due to the reduction in the amount of remote-handled waste are discussed. If DOE should exercise its option to construct a high-level waste repository concurrently with the construction of the revised design, with both facilities receiving waste in 1985, the combined cost would be about 580 million dollars. However, it is unlikely that significant quantities of high-level waste in a form suitable for geologic disposal would be available until after 1990. (13 figures, 5 tables)

  6. WIPP [Waste Isolation Pilot Plant] test phase plan: Performance assessment

    International Nuclear Information System (INIS)

    1990-04-01

    The U.S. Department of Energy (DOE) is responsible for managing the disposition of transuranic (TRU) wastes resulting from nuclear weapons production activities of the United States. These wastes are currently stored nationwide at several of the DOE's waste generating/storage sites. The goal is to eliminate interim waste storage and achieve environmentally and institutionally acceptable permanent disposal of these TRU wastes. The Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico is being considered as a disposal facility for these TRU wastes. This document describes the first of the following two major programs planned for the Test Phase of WIPP: Performance Assessment -- determination of the long-term performance of the WIPP disposal system in accordance with the requirements of the EPA Standard; and Operations Demonstration -- evaluation of the safety and effectiveness of the DOE TRU waste management system's ability to emplace design throughput quantities of TRU waste in the WIPP underground facility. 120 refs., 19 figs., 8 tabs

  7. Environmental Impact Statement: Waste Isolation Pilot Plant (WIPP): Executive summary

    International Nuclear Information System (INIS)

    1980-10-01

    The purpose of this document is to provide a summary of the environmental impact statement for the Waste Isolation Pilot Plant (WIPP) project. The Draft Environmental Impact Statement for the WIPP was published by the US Department of Energy (DOE) in April 1979. This document was reviewed and commented on by members of the general public, private organizations, and governmental agencies. The Final Environmental Impact Statement was subsequently published in October, 1980. This summary is designed to assist decision-maker and interested individuals in reviewing the material presented in the environmental impact statement for the WIPP project. To make this material widely available, this summary is published in both Spanish and English. Additional, more detailed information concerning the environmental and safety consequences of the WIPP project is available in the Final Environmental Impact Statement. Written comments and public hearing comments on the Draft Environmental Impact Statement are available for review. 27 refs., 4 figs., 7 tabs

  8. Final environmental impact statement. Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    This volume contains the appendices for the Final Environmental Impact Statement for the Waste Isolation Pilot Plant (WIPP). Alternative geologic environs are considered. Salt, crystalline rock, argillaceous rock, and tuff are discussed. Studies on alternate geologic regions for the siting of WIPP are reviewed. President Carter's message to Congress on the management of radioactive wastes and the findings and recommendations of the interagency review group on nuclear waste management are included. Selection criteria for the WIPP site including geologic, hydrologic, tectonic, physicochemical compatability, and socio-economic factors are presented. A description of the waste types and the waste processing procedures are given. Methods used to calculate radiation doses from radionuclide releases during operation are presented. A complete description of the Los Medanos site, including archaeological and historic aspects is included. Environmental monitoring programs and long-term safety analysis program are described. (DMC)

  9. Waste Isolation Pilot Plant in situ experimental program for HLW

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1977-01-01

    The Waste Isolation Pilot Plant (WIPP) will be a facility to demonstrate the environmental and operational safety of storing radioactive wastes in a deep geologic bedded salt facility. The WIPP will be located in southeastern New Mexico, approximately 30 miles east of the city of Carlsbad. The major focus of the pilot plant operation involves ERDA defense related low and intermediate-level transuranic wastes. The scope of the project also specifically includes experimentation utilizing commercially generated high-level wastes, or alternatively, spent unreprocessed fuel elements. WIPP HLW experiments are being conducted in an inter-related laboratory, bench-scale, and in situ mode. This presentation focuses on the planned in situ experiments which, depending on the availability of commercially reprocessed waste plus delays in the construction schedule of the WIPP, will begin in approximately 1985. Such experiments are necessary to validate preceding laboratory results and to provide actual, total conditions of geologic storage which cannot be adequately simulated. One set of planned experiments involves emplacing bare HLW fragments into direct contact with the bedded salt environment. A second set utilizes full-size canisters of waste emplaced in the salt in the same manner as planned for a future HLW repository. The bare waste experiments will study in an accelerated manner waste-salt bed-brine interactions including matrix integrity/degradation, brine leaching, system chemistry, and potential radionuclide migration through the salt bed. Utilization of full-size canisters of HLW in situ permits us to demonstrate operational effectiveness and safety. Experiments will evaluate corrosion and compatibility interactions between the waste matrix, canister and overpack materials, getter materials, stored energy, waste buoyancy, etc. Using full size canisters also allows us to demonstrate engineered retrievability of wastes, if necessary, at the end of experimentation

  10. Test phase plan for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1993-03-01

    The US Department of Energy (DOE) has prepared this Test Phase Plan for the Waste Isolation Pilot Plant to satisfy the requirements of Public Law 102-579, the Waste Isolation Pilot Plant (WIPP) Land Withdrawal Act (LWA). The Act provides seven months after its enactment for the DOE to submit this Plan to the Environmental Protection Agency (EPA) for review. A potential geologic repository for transuranic wastes, including transuranic mixed wastes, generated in national-defense activities, the WIPP is being constructed in southeastern New Mexico. Because these wastes remain radioactive and chemically hazardous for a very long time, the WIPP must provide safe disposal for thousands of years. The DOE is developing the facility in phases. Surface facilities for receiving waste have been built and considerable underground excavations (2150 feet below the surface) that are appropriate for in-situ testing, have been completed. Additional excavations will be completed when they are required for waste disposal. The next step is to conduct a test phase. The purpose of the test phase is to develop pertinent information and assess whether the disposal of transuranic waste and transuranic mixed waste in the planned WIPP repository can be conducted in compliance with the environmental standards for disposal and with the Solid Waste Disposal Act (SWDA) (as amended by RCRA, 42 USC. 6901 et. seq.). The test phase includes laboratory experiments and underground tests using contact-handled transuranic waste. Waste-related tests at WIPP will be limited to contact-handled transuranic and simulated wastes since the LWA prohibits the transport to or emplacement of remote-handled transuranic waste at WIPP during the test phase

  11. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    Energy Technology Data Exchange (ETDEWEB)

    Westinghouse TRU Solutions

    2000-12-01

    This Biennial Environmental Compliance Report (BECR) documents environmental regulatory compliance at the Waste Isolation Pilot Plant (WIPP), a facility designed for the safe disposal of transuranic (TRU) radioactive waste, for the reporting period of April 1, 1998, to March 31, 2000. As required by the WIPP Land Withdrawal Act (LWA)(Public Law [Pub. L.] 102-579, and amended by Pub. L. 104-201), the BECR documents U.S. Department of Energy (DOE) Carlsbad Area Office's (hereinafter the ''CAO'') compliance with applicable environmental protection laws and regulations implemented by agencies of the federal government and the state of New Mexico. An issue was identified in the 1998 BECR relating to a potential cross-connection between the fire-water systems and the site domestic water system. While the CAO and its managing and operating contractor (hereinafter the ''MOC'') believe the site was always in compliance with cross-connection control requirements, hardware and procedural upgrades w ere implemented in March 1999 to strengthen its compliance posture. Further discussion of this issue is presented in section 30.2.2 herein. During this reporting period WIPP received two letters and a compliance order alleging violation of certain requirements outlined in section 9(a)(1) of the LWA. With the exception of one item, pending a final decision by the New Mexico Environment Department (NMED), all alleged violations have been resolved without the assessment of fines or penalties. Non-mixed TRU waste shipments began on March 26, 1999. Shipments continued through November 26, 1999, the effective date of the Waste Isolation Pilot Plant Hazardous Waste Facility Permit (NM4890139088-TSDF). No shipments regulated under the Hazardous Waste Facility Permit were received at WIPP during this BECR reporting period.

  12. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    International Nuclear Information System (INIS)

    Westinghouse TRU Solutions

    2000-01-01

    This Biennial Environmental Compliance Report (BECR) documents environmental regulatory compliance at the Waste Isolation Pilot Plant (WIPP), a facility designed for the safe disposal of transuranic (TRU) radioactive waste, for the reporting period of April 1, 1998, to March 31, 2000. As required by the WIPP Land Withdrawal Act (LWA)(Public Law [Pub. L.] 102-579, and amended by Pub. L. 104-201), the BECR documents U.S. Department of Energy (DOE) Carlsbad Area Office's (hereinafter the ''CAO'') compliance with applicable environmental protection laws and regulations implemented by agencies of the federal government and the state of New Mexico. An issue was identified in the 1998 BECR relating to a potential cross-connection between the fire-water systems and the site domestic water system. While the CAO and its managing and operating contractor (hereinafter the ''MOC'') believe the site was always in compliance with cross-connection control requirements, hardware and procedural upgrades w ere implemented in March 1999 to strengthen its compliance posture. Further discussion of this issue is presented in section 30.2.2 herein. During this reporting period WIPP received two letters and a compliance order alleging violation of certain requirements outlined in section 9(a)(1) of the LWA. With the exception of one item, pending a final decision by the New Mexico Environment Department (NMED), all alleged violations have been resolved without the assessment of fines or penalties. Non-mixed TRU waste shipments began on March 26, 1999. Shipments continued through November 26, 1999, the effective date of the Waste Isolation Pilot Plant Hazardous Waste Facility Permit (NM4890139088-TSDF). No shipments regulated under the Hazardous Waste Facility Permit were received at WIPP during this BECR reporting period

  13. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  14. Failure cause and failure rate evaluation on pumps of BWR plants in PSA. Hypothesis testing for typical or plant specific failure rate of pumps

    International Nuclear Information System (INIS)

    Sanada, Takahiro; Nakamura, Makoto

    2009-01-01

    In support of domestic nuclear industry effort to gather and analyze failure data of components concerning nuclear power plants, Nuclear Information Archives (NUCIA) are published for useful information to help PSA. This report focuses on NUCIA pertaining to pumps in domestic nuclear power plants, and provides the reliable estimation on failure rate of pumps resulting from failure cause analysis and hypothesis testing of classified and plant specific failure rate of pumps for improving quality in PSA. The classified and plant specific failure rate of pumps are estimated by analyzing individual domestic nuclear power plant's data of 26 Boiling Water Reactors (BWRs) concerning functionally structurally classified pump failures reported from beginning of commercial operation to March 31, 2007. (author)

  15. A pilot plant for solar-cell manufacture; Ligne pilote de fabrication de cellules solaires

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, D.; Ziegler, Y.; Closset, A. [VHF - Technologies SA, Yverdon-les-Bains (Switzerland)

    2005-07-01

    A pilot plant for the manufacture of amorphous silicon solar cells on plastic film substrate was built allowing the annual production of 40 kW peak power. The production steps comprise: a) the continuous coating of n-i-p solar cells by VHF-PECVD with a capacity of 28.5 meters in 8.5 hours; b) transparent-conducting-oxide (TCO) top contact structuring using a continuous process; c) series connection step (scribing and Ag-paste) with a capacity of 28 meters in 6 hours; d) back and top contact sputtering with 3 parallel magnetrons; e) integration of a large-area vacuum laminator enabling the simultaneous lamination of 4 products of 4 Wp. In parallel with this project, a complete cost model was established enabling a more quantitative approach of the future technological and industrial strategy of the company. An increase of the capacity to 100 kWp has been planned for summer 2005.

  16. The working of RVNRL pilot plant of Rubber Board and it's safety devices

    International Nuclear Information System (INIS)

    Britto, I.J.; Thomas, E.V.

    1996-01-01

    A pilot plant for producing radiation vulcanized natural rubber latex (RVNRL) was established at Rubber Board, India in 1992. Irradiation is done by a batch process in the plant. The plant has a versatile safety system for safety of operators and people working in and around the plant

  17. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Nei, Hiromichi; Hashiguchi, Isao; Inai, Nobuhiko.

    1996-01-01

    A heat exchanger is disposed between a reactor pressure vessel and a turbine, an inlet of a primary circuit of the heat exchanger is connected to a steam pipeline, an exit of the primary circuit of the heat exchanger is connected to a primary coolant pipeline, the primary coolant pipeline is connected to a feed water pipeline, a secondary circuit steam pipeline connected to the heat exchanger is connected to the turbine and a condensate circuit from the turbine is connected to the secondary coolant pipeline connected to the heat exchanger. Steams generated in the reactor are once flown into the heat exchanger to heat secondary coolants indirectly in the heat exchanger, and the generated steams are introduced to the steam turbine. Incondensible gases generated from the reactor and inflowing to the primary side of the heat exchanger are introduced, together with a portion of the steams, to a small-sized condensator passing through steam pipelines in the vicinity of a water surface in a hot well for storing condensed water and disposed at the lower portion of the heat exchanger, the steams and the incondensible gases are separated, and the incondensible gases are processed in an extraction system. Then, steam condition is improved to an over-heat state, and no large-sized shieldings are necessary. (N.H.)

  18. BWR nuclear plant maintenance simulation

    International Nuclear Information System (INIS)

    Stuart, I.F.

    1985-01-01

    As early as 1977, the General Electric Company, USA, Nuclear Energy Operation was making plans to construct a maintenance-type simulator to support Training and Services. The Company's pioneering experience with control room simulators started in 1968 with the Dresden simulator and showed clearly the benefits of having such facilities for training, checkout of procedures and, in the case of maintenance, match-up of equipment or tools as needed. Since the dedication of the facility, it has proved to be an invaluable resource in the training of refuelling and servicing crews. The facility has also been extensively used as developmental and test facility for in-vessel servicing equipment and procedures. (author)

  19. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  20. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  1. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  2. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  3. The pilot plant for electron beam food processing

    International Nuclear Information System (INIS)

    Migdal, W.; Stachowicz, W.

    1993-01-01

    The investigations on food irradiation began in Poland in the end of 50-ties. Till the end of 70-ties the research activity on food irradiation was rather of the random nature and the objectives involved the fundamental research areas of food science. After the JECFI recommended in 1980 the general approval of foods treated with the doses of ionizing radiation up to 10 kG as unconditionally wholesome, the interest on practical application of food irradiation was gained in Poland. In 1986 the governmental bodies decided to recognize the possibilities of practical application of radiation techniques in agriculture, and the Central Research and Development Project No 10.13. ''Radiation Techniques in Agriculture'' was initiated for the period of 5 years. The project in the part that refers to food irradiations involved 3 major objectives: - radiation preservation of food; - radiation hygienization of animal feed; - Pilot plants for food irradiation. The most liable project of the programme was the construction of experimental plant for electron beam food irradiation, intended to be the national center for future testing and implementary works in this field. (orig.)

  4. Preliminary design needs for pilot plant of Monazite processing into Thorium Oxide (ThO_2)

    International Nuclear Information System (INIS)

    Hafni Lissa Nuri; Prayitno; Abdul Jami; M-Pancoko

    2014-01-01

    Data and information collection aimed in order to meet the needs of the initial design for pilot plant of monazite processing into thorium oxide (ThO_2). The content of thorium in monazite is high in Indonesia between 2.9 to 4.1% and relatively abundant in Bangka Belitung Islands. Thorium can be used as fuel because of its potential is more abundant instead of uranium. Plant of thorium oxide commercially from monazite established starting from pilot uranium. Plant of thorium oxide commercially from monazite established starting from pilot plant in order to test laboratory data. Pilot plant design started from initial design, basic design, detailed design, procurement and construction. Preliminary design needs includes data feed and products, a block diagram of the process, a description of the process, the determination of process conditions and type of major appliance has been conducted. (author)

  5. General Atomic Reprocessing Pilot Plant: engineering-scale dissolution system description

    International Nuclear Information System (INIS)

    Yip, H.H.

    1979-04-01

    In February 1978, a dissolver-centrifuge system was added to the cold reprocessing pilot plant at General Atomic Company, which completed the installation of an HTGR fuel head-end reprocessing pilot plant. This report describes the engineering-scale equipment in the pilot plant and summarizes the design features derived from development work performed in the last few years. The dissolver operating cycles for both thorium containing BISO and uranium containinng WAR fissile fuels are included. A continuous vertical centrifuge is used to clarify the resultant dissolver product solution. Process instrumentation and controls for the system reflect design philosophy suitable for remote operation

  6. The UCOR pilot plant and the development of axial flow compressors

    International Nuclear Information System (INIS)

    Grant, W.

    1984-01-01

    This article discusses some of the mechanical aspects of the Uranium Enrichment Corporation of South Africa (Pty) Ltd. (UCOR) pilot plant. The most important mechanical components in a typical stage in the pilot plant, consists of a compressor which is used to compress the process gas mixture. After air cooling to almost room temperature, the mixture is fed through the separation elements. Other components are two pressure vessels connected to the compressor. The development and characteristics of the pilot plant is described in the article

  7. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  8. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  9. Waste Isolation Pilot Plant Annual Site Enviromental Report for 2008

    International Nuclear Information System (INIS)

    2009-01-01

    The purpose of the Waste Isolation Pilot Plant Annual Site Environmental Report for 2008 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to characterize site environmental management performance; summarize environmental occurrences and responses reported during the calendar year; confirm compliance with environmental standards and requirements; highlight significant facility programs and efforts; and describe how compliance and environmental improvement is accomplished through the WIPP Environmental Management System (EMS). The DOE Carlsbad Field Office (CBFO) and the management and operating contractor (MOC), Washington TRU Solutions LLC (WTS), maintain and preserve the environmental resources at the Waste Isolation Pilot Plant (WIPP). DOE Order 231.1A; DOE Order 450.1A, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and the Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and workers, and preservation of the environment. This report was prepared in accordance with DOE Order 231.1A, which requires that DOE facilities submit an ASER to the DOE Headquarters Chief Health, Safety, and Security Officer. The WIPP Hazardous Waste Facility Permit (HWFP) Number NM4890139088-TSDF (treatment, storage, and disposal facility) further requires that the ASER be provided to the New Mexico Environment Department (NMED). The WIPP mission is to safely dispose of transuranic (TRU) radioactive waste generated by the production of nuclear weapons and other activities related to the national defense of the United States. In 2008, 5,265 cubic meters (m3) of TRU waste were disposed of at the WIPP facility, including 5,216 m3 of contact-handled (CH) TRU waste and 49 m3 of remote-handled (RH) TRU waste. From the first

  10. Waste Isolation Pilot Plant Annual Site Enviromental Report for 2008

    Energy Technology Data Exchange (ETDEWEB)

    Washington Regulatory and Enviromnetal Services

    2009-09-21

    The purpose of the Waste Isolation Pilot Plant Annual Site Environmental Report for 2008 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to characterize site environmental management performance; summarize environmental occurrences and responses reported during the calendar year; confirm compliance with environmental standards and requirements; highlight significant facility programs and efforts; and describe how compliance and environmental improvement is accomplished through the WIPP Environmental Management System (EMS). The DOE Carlsbad Field Office (CBFO) and the management and operating contractor (MOC), Washington TRU Solutions LLC (WTS), maintain and preserve the environmental resources at the Waste Isolation Pilot Plant (WIPP). DOE Order 231.1A; DOE Order 450.1A, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and the Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and workers, and preservation of the environment. This report was prepared in accordance with DOE Order 231.1A, which requires that DOE facilities submit an ASER to the DOE Headquarters Chief Health, Safety, and Security Officer. The WIPP Hazardous Waste Facility Permit (HWFP) Number NM4890139088-TSDF (treatment, storage, and disposal facility) further requires that the ASER be provided to the New Mexico Environment Department (NMED). The WIPP mission is to safely dispose of transuranic (TRU) radioactive waste generated by the production of nuclear weapons and other activities related to the national defense of the United States. In 2008, 5,265 cubic meters (m3) of TRU waste were disposed of at the WIPP facility, including 5,216 m3 of contact-handled (CH) TRU waste and 49 m3 of remote-handled (RH) TRU waste. From the first

  11. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  12. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

  13. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  14. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

  15. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  16. Seismic reflection data report: Waste Isolation Pilot Plant (WIPP) site, Southeastern New Mexico

    International Nuclear Information System (INIS)

    Hern, J.L.; Powers, D.W.; Barrows, L.J.

    1978-12-01

    Volume II contains uninterpreted processed lines and shotpoint maps from three seismic reflection surveys conducted from 1976 through 1978 by Sandia Laboratories to support investigations for the Waste Isolation Pilot Plant. Data interpretations will be the subject of subsequent reports

  17. Study on designing a complete pilot plant for processing sandstone ores in Palua-Parong area

    International Nuclear Information System (INIS)

    Le Quang Thai; Tran Van Son; Tran The Dinh; Trinh Nguyen Quynh; Vu Khac Tuan

    2015-01-01

    Design work is the first step of the construction and operation of pilot plant. Thus, the project Study on designing a complete pilot plant for processing sandstone ores in Palua - Parong area was conducted to design a pilot plant for testing entire technological process to obtain yellowcake. Based on a literature review of uranium ore processing technology in the world, information of ore and previous research results of uranium ore in PaLua - PaRong area at the Institute for Technology of Radioactive and Rare Elements, a suitable technological flowsheet for processing this ore has been selected. The size, location of the pilot plant and planed experiments has been selected during the implementation of this project, in which basic parameters, designed system of equipment, buildings, ect. were also calculated. (author)

  18. NPDES Permit for Crow Municipal Rural & Industrial Pilot Water Treatment Plant in Montana

    Science.gov (United States)

    Under NPDES permit MT-0031827, the Crow Indian Tribe is authorized to discharge from the Crow Municipal Rural & Industrial (MR&I) Pilot Water Treatment Plant in Bighorn County, Montana to the Bighorn River.

  19. WIPP conceptual design report. Addendum C. Cost worksheets for Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    1977-04-01

    The cost worksheets for the Waste Isolation Pilot Plant (WIPP) are presented. A summary cost estimate, cost estimate for surface facilities, and cost estimate for shafts and underground facilities are included

  20. WIPP conceptual design report. Addendum G. Accident analysis for Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Shefelbine, H.C.; Metcalf, J.H.

    1977-06-01

    The types of accidents or risks pertinent to the Waste Isolation Pilot Plant (WIPP) are presented. Design features addressing these risks are discussed. Also discussed are design features that protect the public

  1. Waste Isolation Pilot Plant No-migration variance petition. Addendum: Volume 7, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1990-03-01

    This report describes various aspects of the Waste Isolation Pilot Plant (WIPP) including design data, waste characterization, dissolution features, ground water hydrology, natural resources, monitoring, general geology, and the gas generation/test program.

  2. NOx Abatement Pilot Plant 90-day test results report

    International Nuclear Information System (INIS)

    McCray, J.A.; Boardman, R.D.

    1991-01-01

    High-level radioactive liquid wastes produced during nuclear fuel reprocessing at the Idaho Chemical Processing Plant are calcined in the New Waste Calcining Facility (NWCF) to provide both volume reduction and a more stable waste form. Because a large component of the HLW is nitric acid, high levels of oxides of nitrogen (NO x ) are produced in the process and discharged to the environment via the calciner off-gas. The NO x abatement program is required by the new Fuel Processing Restoration (FPR) project permit to construct to reduce NO x emissions from the NWCF. Extensive research and development has indicated that the selective catalytic reduction (SCR) process is the most promising technology for treating the NWCF off-gas. Pilot plant tests were performed to determine the compatibility of the SCR process with actual NWCF off-gas. Test results indicate that the SCR process is a viable method for abating the NO x from the NWCF off-gas. Reduction efficiencies over 95% can be obtained, with minimal amounts of ammonia slip, provided favorable operating conditions exist. Two reactors operated with series flow will provide optimum reduction capabilities. Typical operation should be performed with a first reactor stage gas space velocity of 20,000 hr -1 and an inlet temperature of 320 degrees C. The first stage exhaust NO x concentration will then dictate the parameter settings for the second stage. Operation should always strive for a peak reactor temperature of 520 degrees C in both reactors, with minimal NH 3 slip from the second reactor. Frequent fluctuations in the NWCF off-gas NO x concentration will require a full-scale reduction facility that is versatile and quick-responding. Sudden changes in NWCF off-gas NO x concentrations will require quick detection and immediate response to avoid reactor bed over-heating and/or excessive ammonia slip

  3. Mechanical compaction of Waste Isolation Pilot Plant simulated waste

    International Nuclear Information System (INIS)

    Butcher, B.M.; Thompson, T.W.; VanBuskirk, R.G.; Patti, N.C.

    1991-06-01

    The investigation described in this report acquired experimental information about how materials simulating transuranic (TRU) waste compact under axial compressive stress, and used these data to define a model for use in the Waste Isolation Pilot Plant (WIPP) disposal room analyses. The first step was to determine compaction curves for various simultant materials characteristic of TRU waste. Stress-volume compaction curves for various combinations of these materials were than derived to represent the combustible, metallic, and sludge waste categories. Prediction of compaction response in this manner is considered essential for the WIPP program because of the difficulties inherent in working with real (radioactive) waste. Next, full-sized 55-gallon drums of simulated combustible, metallic, and sludge waste were axially compacted. These results provided data that can be directly applied to room consolidation and data for comparison with the predictions obtained in Part 1 of the investigation. Compaction curves, which represent the combustible, metallic, and sludge waste categories, were determined, and a curve for the averaged waste inventory of the entire repository was derived. 9 refs., 31 figs., 12 tabs

  4. The Waste Isolation Pilot Plant Performance Assessment Program

    International Nuclear Information System (INIS)

    Myers, J.; Coons, W.E.; Eastmond, R.; Morse, J.; Chakrabarti, S.; Zurkoff, J.; Colton, I.D.; Banz, I.

    1986-01-01

    The Waste Isolation Pilot Plant (WIPP) Performance Assessment Program involves a comprehensive analysis of the WIPP project with respect to the recently finalized Environmental Protection Agency regulations regarding the long-term geologic isolation of radioactive wastes. The performance assessment brings together the results of site characterization, underground experimental, and environmental studies into a rigorous determination of the performance of WIPP as a disposal system for transuranic radioactive waste. The Program consists of scenario development, geochemical, hydrologic, and thermomechanical support analyses and will address the specific containment and individual protection requirements specified in 40 CFR 191 sub-part B. Calculated releases from these interrelated analyses will be reported as an overall probability distribution of cumulative release resulting from all processes and events occurring over the 10,000 year post-closure period. In addition, results will include any doses to the public resulting from natural processes occurring over the 1,000 year post-closure period. The overall plan for the WIPP Performance Assessment Program is presented along with approaches to issues specific to the WIPP project

  5. Experimental program plan for the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    The US Department of Energy has prepared this Experimental Program Plan for the Waste Isolation Pilot Plant (EPP) to provide a summary of the DOE experimental efforts needed for the performance assessment process for the WIPP, and of the linkages of this process to the appropriate regulations. The Plan encompasses a program of analyses of the performance of the planned repository based on scientific studies, including tests with transuranic waste at laboratory sites, directed at evaluating compliance with the principal regulations governing the WIPP. The Plan begins with background information on the WIPP project, the requirements of the LWA (Land Withdrawal Act), and its objective and scope. It then presents an overview of the regulatory requirements and the compliance approach. Next are comprehensive discussions of plans for compliance with disposal regulations, followed by the SWDA (Solid Waste Disposal Act) and descriptions of activity programs designed to provide information needed for determining compliance. Descriptions and justifications of all currently planned studies designed to support regulatory compliance activities are also included.

  6. Waset Isolation Pilot Plant Annual Site Environmental Report for 2006

    International Nuclear Information System (INIS)

    2007-01-01

    The purpose of the Waste Isolation Pilot Plant Annual Site Environmental Report for 2006 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data that: (a) Characterize site environmental management performance; (b) Summarize environmental occurrences and responses reported during the calendar year; (c) Confirm compliance with environmental standards and requirements; and (d) Highlight significant facility programs and efforts. The DOE Carlsbad Field Office (CBFO) and Washington TRU Solutions LLC (WTS) maintain and preserve the environmental resources at the WIPP site. DOE Order 231.1A; DOE Order 450.1, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This report was prepared in accordance with DOE Order 231.1A. This order requires that DOE facilities submit an ASER to the DOE Headquarters Office of the Assistant Secretary for Environment, Safety, and Health. The WIPP Hazardous Waste Facility Permit (HWFP) (No. NM4890139088-TSDF [treatment, storage, and disposal facility]) further requires that the ASER be provided to the New Mexico Environment Department (NMED).

  7. Regulatory basis for the Waste Isolation Pilot Plant performance assessment

    International Nuclear Information System (INIS)

    Howard, Bryan A.; Crawford, M.B.; Galson, D.A.; Marietta, Melvin G.

    2000-01-01

    The Waste Isolation Pilot Plant (WIPP) is the first operational repository designed for the safe disposal of transuranic (TRU) radioactive waste from the defense programs of the US Department of Energy (DOE). The US Environmental Protection Agency (EPA) is responsible for certifications and regulation of the WIPP facility for the radioactive components of the waste. The EPA has promulgated general radioactive waste disposal standards at 40 CFR Part 191. and WIPP-specific criteria to implement and interpret the generic disposal standards at 40 CFR Part 194. In October 1996. the DOE submitted its Compliance Certification Application (CCA) to the EPA to demonstrate compliance with the disposal standards at Subparts B and C of 40 CFR Part 191. This paper summarizes the development of the overall legal framework for radioactive waste disposal at the WIPP, the parallel development of the WIPP performance assessment (PA), and how the EPA disposal standards and implementing criteria formed the basis for the CCA WIPP PA. The CCA resulted in a certification in May 1998 by the EPA of the WIPP'S compliance with the EPA's disposal standard, thus enabling the WIPP to begin radioactive waste disposal

  8. Key Geomechanics Issues at the Waste Isolation Pilot Plant Geomechanics

    International Nuclear Information System (INIS)

    HANSEN, FRANCIS D.

    1999-01-01

    Mechanical and hydrological properties of rock salt provide excellent bases for geological isolation of hazardous materials. Regulatory compliance determinations for the Waste Isolation Pilot Plant (WIPP) stand as testament to the widely held conclusion that salt provides excellent isolation properties. The WIPP saga began in the 1950s when the U.S. National Academy of Sciences (NAS) recommended a salt vault as a promising solution to the national problem of nuclear waste disposal. For over 20 years, the Scientific basis for the NAS recommendation has been fortified by Sandia National Laboratories through a series of large scale field tests and laboratory investigations of salt properties. These scientific investigations helped develop a comprehensive understanding of salt's 4 reformational behavior over an applicable range of stresses and temperatures. Sophisticated constitutive modeling, validated through underground testing, provides the computational ability to model long-term behavior of repository configurations. In concert with advancement of the mechanical models, fluid flow measurements showed not only that the evaporite lithology was essentially impermeable but that the WIPP setting was hydrologically inactive. Favorable mechanical properties ensure isolation of materials placed in a salt geological setting. Key areas of the geomechanics investigations leading to the certification of WIPP are in situ experiments, laboratory tests, and shaft seal design

  9. Summary of scientific investigations for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Weart, W.D.

    1996-01-01

    The scientific issues concerning disposal of radioactive wastes in salt formations have received 40 years of attention since the National Academy of Sciences (NAS) first addressed this issue in the mid-50s. For the last 21 years, Sandia National Laboratories (SNL) have directed site specific studies for the Waste Isolation Pilot Plant (WIPP). This paper will focus primarily on the WIPP scientific studies now in their concluding stages, the major scientific controversies regarding the site, and some of the surprises encountered during the course of these scientific investigations. The WIPP project's present understanding of the scientific processes involved continues to support the site as a satisfactory, safe location for the disposal of defense-related transuranic waste and one which will be shown to be in compliance with Environmental Protection Agency (EPA) standards. Compliance will be evaluated by incorporating data from these experiments into Performance Assessment (PA) models developed to describe the physical and chemical processes that could occur at the WIPP during the next 10,000 years under a variety of scenarios. The resulting compliance document is scheduled to be presented to the EPA in October 1996 and all relevant information from scientific studies will be included in this application and the supporting analyses. Studies supporting this compliance application conclude the major period of scientific investigation for the WIPP. Further studies will be of a ''confirmatory'' and monitoring nature

  10. Waset Isolation Pilot Plant Annual Site Environmental Report for 2006

    Energy Technology Data Exchange (ETDEWEB)

    Washington Regulatory and Environmental Services; Washington TRU Solutions LLC

    2007-09-26

    The purpose of the Waste Isolation Pilot Plant Annual Site Environmental Report for 2006 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data that: (a) Characterize site environmental management performance; (b) Summarize environmental occurrences and responses reported during the calendar year; (c) Confirm compliance with environmental standards and requirements; and (d) Highlight significant facility programs and efforts. The DOE Carlsbad Field Office (CBFO) and Washington TRU Solutions LLC (WTS) maintain and preserve the environmental resources at the WIPP site. DOE Order 231.1A; DOE Order 450.1, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and the environment. This report was prepared in accordance with DOE Order 231.1A. This order requires that DOE facilities submit an ASER to the DOE Headquarters Office of the Assistant Secretary for Environment, Safety, and Health. The WIPP Hazardous Waste Facility Permit (HWFP) (No. NM4890139088-TSDF [treatment, storage, and disposal facility]) further requires that the ASER be provided to the New Mexico Environment Department (NMED).

  11. Waste Isolation Pilot Plant: Draft Supplement Environmental Impact Statement

    International Nuclear Information System (INIS)

    1989-04-01

    The US Department of Energy (DOE) has prepared this supplement to the 1980 Final Environmental Impact Statement (FEIS) for the Waste Isolation Pilot Plant (WIPP) in order to assess the environmental impacts that may occur from the continued development of the WIPP as a minced geologic repository for transuranic (TRU) waste. Since the publication of the FEIS in October 1980, new data collected at the WIPP have led to changes in the understanding of the hydrogeologic characteristics of the area and their potential implications for the long-term performance of the WIPP. In addition, there have been changes in the FEIS Proposed Action and new regulatory requirements. This supplement to the FEIS (SEIS) evaluates the environmental consequences of the Proposed Action as modified since 1980 in light of new data and assumptions. The new information pertains mainly to the geologic and hydrologic systems at the WIPP site and their effect on the long-term performance of the WIPP. The SEIS includes new data indicating that: the permeability of the Salado Formation, the geologic formation in which the WIPP underground facilities are located, is lower than previously believed; the moisture content of the Salado Formation and the consequent brine inflow is higher than previously believed; a higher transmissivity zone is present in the Rustler Formation in the southeastern portion of the WIPP site; and ''salt creep'' (convergence) in the repository occurs faster than previously believed. Volume 2 contains 11 appendices

  12. Waste Isolation Pilot Plant Annual Site Environmental Report for 2010

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of the Waste Isolation Pilot Plant (WIPP) Annual Site Environmental Report for 2010 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to: (1) Characterize site environmental management performance. (2) Summarize environmental occurrences and responses reported during the calendar year. (3) Confirm compliance with environmental standards and requirements. (4) Highlight significant environmental accomplishments, including progress toward the DOE Environmental Sustainability Goals made through implementation of the WIPP Environmental Management System (EMS). The DOE Carlsbad Field Office (CBFO) and the management and operating contractor (MOC), Washington TRU Solutions LLC (WTS), maintain and preserve the environmental resources at the WIPP. DOE Order 231.1A; DOE Order 450.1A, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and the Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and workers, and preservation of the environment. This report was prepared in accordance with DOE Order 231.1A, which requires that DOE facilities submit an ASER to the DOE Headquarters Chief Health, Safety, and Security Officer. The WIPP Hazardous Waste Facility Permit Number NM4890139088-TSDF (Permit) further requires that the ASER be provided to the New Mexico Environment Department (NMED).

  13. Geomechanical applications for the Waste Isolation Pilot Plant (WIPP) project

    International Nuclear Information System (INIS)

    Matalucci, R.V.; Hunter, T.O.

    1981-01-01

    The Waste Isolation Pilot Plant (WIPP) is a research and development facility in bedded salt addressing the technical issues associated with the demonstration of disposal of radioactive waste from the defense programs of the USA. The geomechanical program includes laboratory experimentation, constitutive model and computer code development, and in-situ experimentation. Various material models, including creep for salt, and techniques for predicting room response under thermal and mechanical loads have been developed and are being applied to experiment and facility designs. A Benchmark II study has been conducted to compare the capabilities of nine structural codes to predict response of underground configuration under ambient temperature and with a thermal load of 7.5 W/m 2 . Parametric studies are being conducted to evaluate optimum room configurations. A series of in situ experiments is the next step towards validating models and predictive techniques. These experiments will be conducted in a facility in southeastern New Mexico mined at a depth of 659 m

  14. Compliance status report for the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    1994-03-31

    The US Department of Energy (DOE) is responsible for the disposition of transuranic (TRU) waste generated through national defense-related activities. Approximately 53,700 m{sup 2} of these wastes have been generated and are currently stored at government defense installations across the country. The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been sited and constructed to meet the criteria established by the scientific and regulatory community for the safe, long-term disposal of TRU and TRU-mixed wastes. This Compliance Status Report (CSR) provides an assessment of the progress of the WIPP Program toward compliance with long-term disposal regulations, set forth in Title 40 CFR 191 (EPA, 1993a), Subparts B and C, and Title 40 CFR {section}268.6 (EPA, 1993b), in order to focus on-going and future experimental and engineering activities. The CSR attempts to identify issues associated with the performance of the WIPP as a long-term repository and to focus on the resolution of these issues. This report will serve as a tool to focus project resources on the areas necessary to ensure complete, accurate, and timely submittal of the compliance application. This document is not intended to constitute a statement of compliance or a demonstration of compliance.

  15. The waste isolation pilot plant: A new regulatory environment

    International Nuclear Information System (INIS)

    Frei, M.W.; Schneider, S.P.; Saris, E.C.; Austin, P.W.

    1993-01-01

    The US Department of Energy (DOE) is ready to embark on a multiyear test program, using radioactive waste, at the Waste Isolation Pilot Plant (WIPP). The WIPP is a deep geologic repository, constructed in ancient salt beds in southeastern New Mexico. It was authorized by Congress in 1979 as a research and development facility to demonstrate safe disposal of the nation's defense transuranic (TRU) waste. Nonradioactive testing in the repository has been under way for several years. The DOE is now ready to begin underground experiments at WIPP with small amounts of TRU waste. Radioactive waste testing in an actual repository environment will reduce uncertainties associated with predictions of long-term repository performance. However, the authority for DOE to begin this new phase of the test program no longer resides within the department. The WIPP is now subject to a new level of regulatory oversight by the Environmental Protection Agency (EPA) and other federal agencies, as set forth by Public Law 102-579, the WIPP Land Withdrawal Act, signed by the President on October 30, 1992. This paper discusses the act's new regulatory requirements for WIPP

  16. The waste isolation pilot plant project: a changing paradigm

    International Nuclear Information System (INIS)

    Sheppard, L.E.; McFadden, M.H.

    1996-01-01

    The Waste Isolation Pilot Plant (WIPP) is a US Department of Energy (DOE) repository that has been developed to demonstrate the safe and permanent isolation of transuranic radioactive wastes in a deep geologic site. It is located in 650 m below the surface in a bedded salt formation, and is designed to hold approximately 175,500 cubic meters of waste. Compliance with the regulations has become the principal focus for the Project. The scientific baseline is an important and integral part of the CCA, as it provides the foundation for conducting total system performance assessment calculations for comparison with applicable standards. The activities required to support the scientific baseline are being pursued in parallel to minimize the time required to collect, analyze, interpret and fully incorporate the results into the CCA. The DOE has shifted its approach to demonstrating compliance with the applicable regulations from a paradigm of a series of broad investigations to a new paradigm of highly focused activities conducted in parallel. The success of this approach will be assessed by the EPA when the application is critically reviewed

  17. Waste Isolation Pilot Plant Annual Site Environmental Report for 2010

    Energy Technology Data Exchange (ETDEWEB)

    None

    2011-09-01

    The purpose of the Waste Isolation Pilot Plant (WIPP) Annual Site Environmental Report for 2010 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to: (1) Characterize site environmental management performance. (2) Summarize environmental occurrences and responses reported during the calendar year. (3) Confirm compliance with environmental standards and requirements. (4) Highlight significant environmental accomplishments, including progress toward the DOE Environmental Sustainability Goals made through implementation of the WIPP Environmental Management System (EMS). The DOE Carlsbad Field Office (CBFO) and the management and operating contractor (MOC), Washington TRU Solutions LLC (WTS), maintain and preserve the environmental resources at the WIPP. DOE Order 231.1A; DOE Order 450.1A, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and the Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and workers, and preservation of the environment. This report was prepared in accordance with DOE Order 231.1A, which requires that DOE facilities submit an ASER to the DOE Headquarters Chief Health, Safety, and Security Officer. The WIPP Hazardous Waste Facility Permit Number NM4890139088-TSDF (Permit) further requires that the ASER be provided to the New Mexico Environment Department (NMED).

  18. Waste acceptance criteria for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1996-04-01

    The Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC), DOE/WIPP-069, was initially developed by a U.S. Department of Energy (DOE) Steering Committee to provide performance requirements to ensure public health and safety as well as the safe handling of transuranic (TRU) waste at the WIPP. This revision updates the criteria and requirements of previous revisions and deletes those which were applicable only to the test phase. The criteria and requirements in this document must be met by participating DOE TRU Waste Generator/Storage Sites (Sites) prior to shipping contact-handled (CH) and remote-handled (RH) TRU waste forms to the WIPP. The WIPP Project will comply with applicable federal and state regulations and requirements, including those in Titles 10, 40, and 49 of the Code of Federal Regulations (CFR). The WAC, DOE/WIPP-069, serves as the primary directive for assuring the safe handling, transportation, and disposal of TRU wastes in the WIPP and for the certification of these wastes. The WAC identifies strict requirements that must be met by participating Sites before these TRU wastes may be shipped for disposal in the WIPP facility. These criteria and requirements will be reviewed and revised as appropriate, based on new technical or regulatory requirements. The WAC is a controlled document. Revised/changed pages will be supplied to all holders of controlled copies

  19. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    International Nuclear Information System (INIS)

    2006-01-01

    This Biennial Environmental Compliance Report (BECR) documents compliance with environmental regulations at the Waste Isolation Pilot Plant (WIPP), a facility designed and authorized for the safe disposal of transuranic (TRU) radioactive waste. This BECR covers the reporting period from April 1, 2004, to March 31, 2006. As required by the WIPP Land Withdrawal Act (LWA) (Public Law [Pub. L.] 102-579, as amended by Pub. L. 104-201), the BECR documents United States (U.S.) Department of Energy (DOE) compliance with regulations and permits issued pursuant to the following: (1) Title 40 Code of Federal Regulations (CFR) Part 191, Subpart A, 'Environmental Standards for Management and Storage'; (2) Clean Air Act (CAA) (42 United States Code [U.S.C.] 7401, et seq.); (3) Solid Waste Disposal Act (SWDA) (42 U.S.C. 6901-6992, et seq.); (4) Safe Drinking Water Act (SDWA) (42 U.S.C. 300f, et seq.); (5) Toxic Substances Control Act (TSCA) (15 U.S.C. 2601, et seq.); (6) Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) (42 U.S.C. 9601, et seq.); and all other federal and state of New Mexico laws pertaining to public health and safety or the environment.

  20. A historical review of Waste Isolation Pilot Plant backfill development

    International Nuclear Information System (INIS)

    Krumhansl, James L.; Molecke, Martin A.; Papenguth, Hans W.; Brush, Laurence H.

    2000-01-01

    Backfills have been part of Sandia National Laboratories' [Sandia's] Waste Isolation Pilot Plant [WIPP] designs for over twenty years. Historically, backfill research at Sandia has depended heavily on the changing mission of the WIPP facility. Early testing considered heat producing, high level, wastes. Bentonite/sand/salt mixtures were evaluated and studies focused on developing materials that would retard brine ingress, sorb radionuclides, and withstand elevated temperatures. The present-day backfill consists of pure MgO [magnesium oxide] in a pelletized form and is directed at treating the relatively low contamination level, non-heat producing, wastes actually being disposed of in the WIPP. Its introduction was motivated by the need to scavenging CO 2 [carbon dioxide] from decaying organic components in the waste. However, other benefits, such as a substantial desiccating capacity, are also being evaluated. The MgO backfill also fulfills a statutory requirement for assurance measures beyond those needed to demonstrate compliance with the US Environmental Protection Agency [EPA] regulatory release limits. However, even without a backfill, the WIPP repository design still operates within EPA regulatory release limits

  1. The 1996 performance assessment for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Anderson, D.R.; Jow, H.N.; Marietta, M.G.; Chu, M.S.Y.; Shephard, L.E.; Helton, J.C.; Basabilvazo, G.

    1998-01-01

    The Waste Isolation Pilot Plant (WIPP) is under development by the US Department of Energy (DOE) for the geologic disposal of transuranic (TRU) waste that has been generated at government defense installations in the United States. The WIPP is located in an area of low population density in southeastern New Mexico. Waste disposal will take place in excavated chambers in a bedded salt formation approximately 655 m below the land surface. This presentation describes a performance assessment (PA) carried out at Sandia National Laboratories (SNL) to support the Compliance Certification Application (CCA) made by the DOE to the US Environmental Protection Agency (EPA) in October, 1996, for the certification of the WIPP for the disposal of TRU waste. Based on the CCA supported by the PA described in this presentation, the EPA has issued a preliminary decision to certify the WIPP for the disposal of TRU waste. At present (April 1998), it appears likely that the WIPP will be in operation by the end of 1998

  2. Experimental program plan for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1994-01-01

    The US Department of Energy has prepared this Experimental Program Plan for the Waste Isolation Pilot Plant (EPP) to provide a summary of the DOE experimental efforts needed for the performance assessment process for the WIPP, and of the linkages of this process to the appropriate regulations. The Plan encompasses a program of analyses of the performance of the planned repository based on scientific studies, including tests with transuranic waste at laboratory sites, directed at evaluating compliance with the principal regulations governing the WIPP. The Plan begins with background information on the WIPP project, the requirements of the LWA (Land Withdrawal Act), and its objective and scope. It then presents an overview of the regulatory requirements and the compliance approach. Next are comprehensive discussions of plans for compliance with disposal regulations, followed by the SWDA (Solid Waste Disposal Act) and descriptions of activity programs designed to provide information needed for determining compliance. Descriptions and justifications of all currently planned studies designed to support regulatory compliance activities are also included

  3. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    Energy Technology Data Exchange (ETDEWEB)

    Washington Regulatory and Environmental Services

    2006-10-12

    This Biennial Environmental Compliance Report (BECR) documents compliance with environmental regulations at the Waste Isolation Pilot Plant (WIPP), a facility designed and authorized for the safe disposal of transuranic (TRU) radioactive waste. This BECR covers the reporting period from April 1, 2004, to March 31, 2006. As required by the WIPP Land Withdrawal Act (LWA) (Public Law [Pub. L.] 102-579, as amended by Pub. L. 104-201), the BECR documents United States (U.S.) Department of Energy (DOE) compliance with regulations and permits issued pursuant to the following: (1) Title 40 Code of Federal Regulations (CFR) Part 191, Subpart A, "Environmental Standards for Management and Storage"; (2) Clean Air Act (CAA) (42 United States Code [U.S.C.] §7401, et seq.); (3) Solid Waste Disposal Act (SWDA) (42 U.S.C. §§6901-6992, et seq.); (4) Safe Drinking Water Act (SDWA) (42 U.S.C. §§300f, et seq.); (5) Toxic Substances Control Act (TSCA) (15 U.S.C. §§2601, et seq.); (6) Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) (42 U.S.C. §§9601, et seq.); and all other federal and state of New Mexico laws pertaining to public health and safety or the environment.

  4. Groundwater monitoring at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Kehrman, R.; Broberg, K.; Tatro, G.; Richardson, R.; Dasczcyszak, W.

    1990-01-01

    This paper discusses the Groundwater Monitoring Program (GPM) being conducted at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The Regulatory and Environmental Programs (REP) section of the Environment, Safety and Health department (ES ampersand H) is responsible for conducting environmental monitoring at the WIPP. Groundwater monitoring is one of the ongoing environmental activities currently taking place. The REP section includes water quality sampling and water level monitoring. The WIPP Project is a research and develop facility designed to demonstrate the safe disposal of defense-generated waste in a geologic repository. Water quality sampling for physical, chemical, and radiological parameters has been an ongoing activity at the WIPP site for the past six years, and will continue through the life of the project. The water quality of a well is sampled while the well is continuously pumped. Serial samples of the pumped water are collected and tested for pH, Eh, temperature, specific gravity, specific conductivity, alkalinity, chlorides, divalent cations, ferrous iron, and total iron. Stabilization of serial sampling parameters determined if a representative sample is being obtained, Representative samples are sent to contract laboratories and analyzed for general chemistry, major cations and anions, and radionuclides. 13 refs., 4 figs., 1 tab

  5. Final Report: RPP-WTP Semi-Integrated Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Duignan, M. R.; Adamson, D. J.; Calloway, T. B.; Fowley, M. D.; Qureshi, Z. H.; Steimke, J. L.; Williams, M. R.; Zamecnik, J. R.

    2005-06-01

    In August 2004 the last of the SIPP task testing ended--a task that formally began with the issuance of the RPP-WTP Test Specification in June 2003. The planning for the task was a major effort in itself and culminated with the input of all stakeholders, DOE, Bechtel National, Inc., Washington Group International, in October 2003 at Hanford, WA (Appendix A). This report documents the activities carried out as a result of that planning. Campaign IV, the fourth and final step towards the Semi-Integrated Pilot Plant (SIPP) task, conducted by the Savannah River National Laboratory (SRNL) at the Savannah River Site, was to take the several recycle streams produced in Campaign III, the third step of the task, and combine them with other simulated recycle and chosen waste streams. (Campaign III was fed recycles from Campaign II, as Campaign II was fed by Campaign I.) The combined stream was processed in a fashion that mimicked the pretreatment operations of the DOE River Protection Project--Waste Treatment and Immobilization Plant (RPP-WTP) with the exception of the Ion Exchange Process. The SIPP task is considered semi-integrated because it only deals with the pretreatment operations of the RPP-WTP. That is, the pilot plant starts by receiving waste from the tank farm and ends when waste is processed to the point of being sent for vitrification. The resulting pretreated LAW and HLW simulants produced by the SIPP were shipped to VSL (Vitreous State Laboratory) and successfully vitrified in pilot WTP melters. Within the SIPP task these steps are referred to as Campaigns and there were four Campaigns in all. Campaign I, which is completely different than other campaigns, subjected a simulant of Hanford Tank 241-AY-102/C-106 (AY102) waste to cross-flow ultrafiltration only and in that process several important recycle streams were produced as a result of washing the simulant and cleaning the cross-flow filter. These streams were fed to subsequent campaigns and that work was

  6. Final Report: RPP-WTP Semi-Integrated Pilot Plant

    International Nuclear Information System (INIS)

    Duignan, M. R.; Adamson, D. J.; Calloway, T. B.; Fowley, M. D.; Qureshi, Z. H.; Steimke, J. L.; Williams, M. R.; Zamecnik, J. R.

    2005-01-01

    In August 2004 the last of the SIPP task testing ended--a task that formally began with the issuance of the RPP-WTP Test Specification in June 2003. The planning for the task was a major effort in itself and culminated with the input of all stakeholders, DOE, Bechtel National, Inc., Washington Group International, in October 2003 at Hanford, WA (Appendix A). This report documents the activities carried out as a result of that planning. Campaign IV, the fourth and final step towards the Semi-Integrated Pilot Plant (SIPP) task, conducted by the Savannah River National Laboratory (SRNL) at the Savannah River Site, was to take the several recycle streams produced in Campaign III, the third step of the task, and combine them with other simulated recycle and chosen waste streams. (Campaign III was fed recycles from Campaign II, as Campaign II was fed by Campaign I.) The combined stream was processed in a fashion that mimicked the pretreatment operations of the DOE River Protection Project--Waste Treatment and Immobilization Plant (RPP-WTP) with the exception of the Ion Exchange Process. The SIPP task is considered semi-integrated because it only deals with the pretreatment operations of the RPP-WTP. That is, the pilot plant starts by receiving waste from the tank farm and ends when waste is processed to the point of being sent for vitrification. The resulting pretreated LAW and HLW simulants produced by the SIPP were shipped to VSL (Vitreous State Laboratory) and successfully vitrified in pilot WTP melters. Within the SIPP task these steps are referred to as Campaigns and there were four Campaigns in all. Campaign I, which is completely different than other campaigns, subjected a simulant of Hanford Tank 241-AY-102/C-106 (AY102) waste to cross-flow ultrafiltration only and in that process several important recycle streams were produced as a result of washing the simulant and cleaning the cross-flow filter. These streams were fed to subsequent campaigns and that work was

  7. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  8. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant

    International Nuclear Information System (INIS)

    Lopez R, A.

    2004-01-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  9. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  10. A soil washing pilot plant for removing petroleum hydrocarbons from contaminated soils

    International Nuclear Information System (INIS)

    Toor, I.A.; Roehrig, G.R.

    1992-01-01

    A soil washing pilot plant was built and tested for its ability to remove petroleum hydrocarbons from certain soils. The ITEX soil washing pilot plant is a trailer mountable mobile unit which has a washing capacity of two tons per hour of contaminated soils. A benchscale study was carried out prior to the fabrication of the pilot plant. The first sample was contaminated with diesel fuel while the second sample was contaminated with crude oil. Various nonionic, cationic and anionic cleaning agents were evaluated for their ability to remove petroleum hydrocarbons from these materials. The nonionic cleaning agents were more successful in cleaning the soils in general. The ultimate surfactant choice was based on several factors including cost, biodegradability, cleaning efficiency and other technical considerations. The soil samples were characterized in terms of their particle size distributions. Commercial diesel fuel was carefully mixed in this sand to prepare a representative sample for the pilot plant study. Two pilot runs were made using this material. A multistage washing study was also conducted in the laboratory which indicates that the contamination level can be reduced to 100 ppm using only four stages. Because the pilot plant washing efficiency is twice as high, it is believed that ultimate contamination levels can be reduced to lower levels using the same number of stages. However, this hypothesis has not been demonstrated to date

  11. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

  12. Selective absorption pilot plant for decontamination of fuel reprocessing plant off-gas

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, M.J.; Eby, R.S.; Huffstetler, V.C.

    1977-10-01

    A fluorocarbon-based selective absorption process for removing krypton-85, carbon-14, and radon-222 from the off-gas of conventional light water and advanced reactor fuel reprocessing plants is being developed at the Oak Ridge Gaseous Diffusion Plant in conjunction with fuel recycle work at the Oak Ridge National Laboratory and at the Savannah River Laboratory. The process is characterized by an especially high tolerance for many other reprocessing plant off-gas components. This report presents detailed drawings and descriptions of the second generation development pilot plant as it has evolved after three years of operation. The test facility is designed on the basis of removing 99% of the feed gas krypton and 99.9% of the carbon and radon, and can handle a nominal 15 scfm (425 slm) of contaminated gas at pressures from 100 to 600 psig (7.0 to 42.2 kg/cm/sup 2/) and temperatures from minus 45 to plus 25/sup 0/F (-43 to -4/sup 0/C). Part of the development program is devoted to identifying flowsheet options and simplifications that lead to an even more economical and reliable process. Two of these applicative flowsheets are discussed.

  13. Final environmental impact statement. Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1980-10-01

    In accordance with the National Environmental Policy Act (NEPA) of 1969, the US Department of Energy (DOE) has prepared this document as environmental input to future decisions regarding the Waste Isolation Pilot Plant (WIPP), which would include the disposal of transuranic waste, as currently authorized. The alternatives covered in this document are the following: (1) Continue storing transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) as it is now or with improved confinement. (2) Proceed with WIPP at the Los Medanos site in southeastern New Mexico, as currently authorized. (3) Dispose of TRU waste in the first available repository for high-level waste. The Los Medanos site would be investigated for its potential suitability as a candidate site. This is administration policy and is the alternative preferred by the DOE. (4) Delay the WIPP to allow other candidate sites to be evaluated for TRU-waste disposal. This environmental impact statement is arranged in the following manner: Chapter 1 is an overall summary of the analysis contained in the document. Chapters 2 and 4 set forth the objectives of the national waste-management program and analyze the full spectrum of reasonable alternatives for meeting these objectives, including the WIPP. Chapter 5 presents the interim waste-acceptance criteria and waste-form alternatives for the WIPP. Chapters 6 through 13 provide a detailed description and environmental analysis of the WIPP repository and its site. Chapter 14 describes the permits and approvals necessary for the WIPP and the interactions that have taken place with Federal, State, and local authorities, and with the general public in connection with the repository. Chapter 15 analyzes the many comments received on the DEIS and tells what has been done in this FEIS in response. The appendices contain data and discussions in support of the material in the text

  14. Waste Isolation Pilot Plant Groundwater Protection Management Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Washington Regulatory and Environmental Services

    2005-07-01

    The DOE established the Groundwater Monitoring Program (GMP) (WP 02-1) to monitor groundwater resources at WIPP. In the past, the GMP was conducted to establish background data of existing conditions of groundwater quality and quantity in the WIPP vicinity, and to develop and maintain a water quality database as required by regulation. Today the GMP is conducted consistent with 204.1.500 NMAC (New MexicoAdministrative Code), "Adoption of 40 CFR [Code of Federal Regulations] Part 264,"specifically 40 CFR §264.90 through §264.101. These sections of 20.4.1 NMAC provide guidance for detection monitoring of groundwater that is, or could be, affected by waste management activities at WIPP. Detection monitoring at WIPP is designed to detect contaminants in the groundwater long before the general population is exposed. Early detection will allow cleanup efforts to be accomplished before any exposure to the general population can occur. Title 40 CFR Part 264, Subpart F, stipulates minimum requirements of Resource Conservation and Recovery Act of 1976 (42 United States Code [U.S.C.] §6901 et seq.) (RCRA) groundwater monitoring programs including the number and location of monitoring wells; sampling and reporting schedules; analytical methods and accuracy requirements; monitoring parameters; and statistical treatment of monitoring data. This document outlines how WIPP intends to protect and preserve groundwater within the WIPP Land Withdrawal Area (WLWA). Groundwater protection is just one aspect of the WIPP environmental protection effort. An overview of the entire environmental protection effort can be found in DOE/WIPP 99-2194, Waste Isolation Pilot Plant Environmental Monitoring Plan. The WIPP GMP is designed to statistically determine if any changes are occurring in groundwater characteristics within and surrounding the WIPP facility. If a change is noted, the cause will then be determined and the appropriate corrective action(s) initiated.

  15. The Waste Isolation Pilot Plant (WIPP) integrated project management system

    International Nuclear Information System (INIS)

    Olona, D.; Sala, D.

    1993-01-01

    The Waste Isolation Pilot Plant (WIPP), located 26 miles east of Carlsbad, New Mexico, is a research and development project of the Department of Energy (DOE), tasked with the mission of demonstrating the safe disposal of transuranic (TRU) radioactive wastes. This unique project was authorized by Congress in 1979 in response to the national need for long-term, safe methods for disposing of radioactive by-products from our national defense programs. The WIPP was originally established in December of 1979, by Public Law 96-164, DOE National Security and Military Applications of Nuclear Energy Authorization Act of 1980. Since the inception of the WIPP Project, work has continued to prepare the facility to receive TRU wastes. Studies continue to be conducted to demonstrate the safety of the WIPP facility in accordance with federal and state laws, state agreements, environmental regulations, and DOE Orders. The objectives of implementing an integrated project management system are to assure compliance with all regulatory and federal regulations, identify areas of concern, provide justification for funding, provide a management tool for control of program workscope, and establish a project baseline from which accountability and performance will be assessed. Program management and project controls are essential for the success of the WIPP Project. The WIPP has developed an integrated project management system to establish the process for the control of the program which has an expected total dollar value of $2B over the ten-year period from 1990-2000. The implementation of this project management system was motivated by the regulatory requirements of the project, the highly public environment in which the project takes place, limited funding and resources, and the dynamic nature of the project. Specific areas to be addressed in this paper include strategic planning, project organization, planning and scheduling, fiscal planning, and project monitoring and reporting

  16. Waste Isolation Pilot Plant Groundwater Protection Management Program Plan

    International Nuclear Information System (INIS)

    2005-01-01

    The DOE established the Groundwater Monitoring Program (GMP) (WP 02-1) to monitor groundwater resources at WIPP. In the past, the GMP was conducted to establish background data of existing conditions of groundwater quality and quantity in the WIPP vicinity, and to develop and maintain a water quality database as required by regulation. Today the GMP is conducted consistent with 204.1.500 NMAC (New Mexico Administrative Code), 'Adoption of 40 CFR [Code of Federal Regulations] Part 264,'specifically 40 CFR 264.90 through 264.101. These sections of 20.4.1 NMAC provide guidance for detection monitoring of groundwater that is, or could be, affected by waste management activities at WIPP. Detection monitoring at WIPP is designed to detect contaminants in the groundwater long before the general population is exposed. Early detection will allow cleanup efforts to be accomplished before any exposure to the general population can occur. Title 40 CFR Part 264, Subpart F, stipulates minimum requirements of Resource Conservation and Recovery Act of 1976 (42 United States Code [U.S.C.] 6901 et seq.) (RCRA) groundwater monitoring programs including the number and location of monitoring wells; sampling and reporting schedules; analytical methods and accuracy requirements; monitoring parameters; and statistical treatment of monitoring data. This document outlines how WIPP intends to protect and preserve groundwater within the WIPP Land Withdrawal Area (WLWA). Groundwater protection is just one aspect of the WIPP environmental protection effort. An overview of the entire environmental protection effort can be found in DOE/WIPP 99-2194, Waste Isolation Pilot Plant Environmental Monitoring Plan. The WIPP GMP is designed to statistically determine if any changes are occurring in groundwater characteristics within and surrounding the WIPP facility. If a change is noted, the cause will then be determined and the appropriate corrective action(s) initiated.

  17. Waste retrieval plan for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1993-03-01

    The US DOE has prepared this plan to meet the requirements of Public Law 102579, the Waste Isolation Pilot Plant (WIPP) LWA, The purpose. is to demonstrate readiness to retrieve from the WIPP underground transuranic radioactive waste that will be used for testing should retrieval be needed. The WIPP, a potential geologic repository for transuranic wastes generated in national-defense activities, has been constructed in southeastern New Mexico. Because the transuranic wastes will remain radioactive for a very long time, the WIPP must reasonably ensure safe performance over thousands of years. The DOE therefore decided to develop the facility in phases, to preclude premature decisions and to conduct the performance assessments needed to demonstrate long-term safety. Surface facilities for receiving waste have been built, and considerable underground excavation, 2150 feet below the surface, has been completed. The next step is a test phase, including underground experiments called ''bin tests'' and ''alcove test(s)'' with contact-handled transuranic waste. The objective of these waste tests is to collect relevant data about the gas-generation potential and volatile organic compound (VOC) source term of the waste for developing a basis for demonstrating long term safety by compliance with the applicable disposal regulations (40 CFR 191, 264 and 268). The test phase will end when a decision is made to begin disposal in the WIPP or to terminate the project if regulatory compliance cannot be determined and demonstrated. Authorization to receive transuranic waste at the WIPP for the test phase is given by the WIPP LWA provided certain requirements are met

  18. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    Energy Technology Data Exchange (ETDEWEB)

    Washinton TRU Solutions LLC

    2002-09-30

    This Biennial Environmental Compliance Report (BECR) documents environmental regulatory compliance at the Waste Isolation Pilot Plant (WIPP), a facility designed for the safe disposal of transuranic (TRU) radioactive waste, for the reporting period of April 1, 2000, to March 31, 2002. As required by the WIPP Land Withdrawal Act (LWA)(Public Law [Pub. L.] 102-579, as amended by Pub. L. 104-201), the BECR documents U.S. Department of Energy (DOE) Carlsbad Field Office's (CBFO) compliance with applicable environmental protection laws and regulations implemented by agencies of the federal government and the state of New Mexico. In the prior BECR, the CBFO and the management and operating contractor (MOC)committed to discuss resolution of a Letter of Violation that had been issued by the New Mexico Environment Department (NMED) in August 1999, which was during the previous BECR reporting period. This Letter of Violation alleged noncompliance with hazardous waste aisle spacing, labeling, a nd tank requirements. At the time of publication of the prior BECR, resolution of the Letter of Violation was pending. On July 7, 2000, the NMED issued a letter noting that the aisle spacing and labeling concerns had been adequately addressed and that they were rescinding the violation alleging that the Exhaust Shaft Catch Basin failed to comply with the requirements for a hazardous waste tank. During the current reporting period, WIPP received a Notice of Violation and a compliance order alleging the violation of the New Mexico Hazardous Waste Regulations and the WIPP Hazardous Waste Facility Permit (HWFP).

  19. Final environmental impact statement. Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    In accordance with the National Environmental Policy Act (NEPA) of 1969, the US Department of Energy (DOE) has prepared this document as environmental input to future decisions regarding the Waste Isolation Pilot Plant (WIPP), which would include the disposal of transuranic waste, as currently authorized. The alternatives covered in this document are the following: (1) Continue storing transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) as it is now or with improved confinement. (2) Proceed with WIPP at the Los Medanos site in southeastern New Mexico, as currently authorized. (3) Dispose of TRU waste in the first available repository for high-level waste. The Los Medanos site would be investigated for its potential suitability as a candidate site. This is administration policy and is the alternative preferred by the DOE. (4) Delay the WIPP to allow other candidate sites to be evaluated for TRU-waste disposal. This environmental impact statement is arranged in the following manner: Chapter 1 is an overall summary of the analysis contained in the document. Chapters 2 and 4 set forth the objectives of the national waste-management program and analyze the full spectrum of reasonable alternatives for meeting these objectives, including the WIPP. Chapter 5 presents the interim waste-acceptance criteria and waste-form alternatives for the WIPP. Chapters 6 through 13 provide a detailed description and environmental analysis of the WIPP repository and its site. Chapter 14 describes the permits and approvals necessary for the WIPP and the interactions that have taken place with Federal, State, and local authorities, and with the general public in connection with the repository. Chapter 15 analyzes the many comments received on the DEIS and tells what has been done in this FEIS in response. The appendices contain data and discussions in support of the material in the text.

  20. Groundwater monitoring at the waste isolation pilot plant

    International Nuclear Information System (INIS)

    Kehrman, R.; Broberg, K.; Tatro, G.; Richardson, R.; Dasczcyszak, W.

    1990-01-01

    This paper discusses the Groundwater Monitoring Program (GMP) being conducted at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The Regulatory and Environmental Programs (REP) section of the Environment, Safety and Health department (ES ampersand H) is responsible for conducting environmental monitoring at the WIPP. Groundwater monitoring is one of the ongoing environmental activities currently taking place. The REP section includes water-quality sampling and water-level monitoring. The WIPP Project is a research and development facility designed to demonstrate the safe disposal of defense-generated TRU and mixed waste in a geologic repository. The Salado Formation of Permian age serves as the repository medium. The Salado Formation consists of bedded salt and associated evaporites. The formation is 602 meters thick at the site area; the top surface is located at a subsurface depth of 262 meters (10). The repository lies at a subsurface depth of 655 meters. Water-quality sampling for physical, chemical, and radiological parameters has been an ongoing activity at the WIPP site for the past six years, and will continue through the life of the project. Data collected from this program to date, has been used by Sandia National Laboratories for site characterization and performance assessment work. The data has also been used to establish a baseline of preoperational radiological and nonradiological groundwater quality. Once the facility begins receiving waste, this baseline will be used to determine if the WIPP facility influences or alters groundwater quality over time. The water quality of a well is determined while the well is continuously pumped. Serial samples of the pumped water are collected and tested for pH, Eh, temperature, specific gravity, specific conductivity, alkalinity, chlorides, divalent cations, ferrous iron, and total iron. 13 refs., 4 figs., 1 tab

  1. Waste Isolation Pilot Plant Biennial Environmental Compliance Report

    International Nuclear Information System (INIS)

    Washinton TRU Solutions LLC

    2002-01-01

    This Biennial Environmental Compliance Report (BECR) documents environmental regulatory compliance at the Waste Isolation Pilot Plant (WIPP), a facility designed for the safe disposal of transuranic (TRU) radioactive waste, for the reporting period of April 1, 2000, to March 31, 2002. As required by the WIPP Land Withdrawal Act (LWA)(Public Law [Pub. L.] 102-579, as amended by Pub. L. 104-201), the BECR documents U.S. Department of Energy (DOE) Carlsbad Field Office's (CBFO) compliance with applicable environmental protection laws and regulations implemented by agencies of the federal government and the state of New Mexico. In the prior BECR, the CBFO and the management and operating contractor (MOC)committed to discuss resolution of a Letter of Violation that had been issued by the New Mexico Environment Department (NMED) in August 1999, which was during the previous BECR reporting period. This Letter of Violation alleged noncompliance with hazardous waste aisle spacing, labeling, a nd tank requirements. At the time of publication of the prior BECR, resolution of the Letter of Violation was pending. On July 7, 2000, the NMED issued a letter noting that the aisle spacing and labeling concerns had been adequately addressed and that they were rescinding the violation alleging that the Exhaust Shaft Catch Basin failed to comply with the requirements for a hazardous waste tank. During the current reporting period, WIPP received a Notice of Violation and a compliance order alleging the violation of the New Mexico Hazardous Waste Regulations and the WIPP Hazardous Waste Facility Permit (HWFP)

  2. The 10 MWe Solar Thermal Central Receiver Pilot Plant: Solar facilities design integration. Pilot-plant station manual (RADL Item 2-1). Volume 1: System description

    Science.gov (United States)

    1982-09-01

    The complete Barstow Solar Pilot Plant is described. The plant requirements and general description are presented, the mechanical, electric power, and control and instrumentation systems as well as civil engineering and structural aspects and the station buildings are described. Included in the mechanical systems are the heliostats, receiver, thermal storage system, beam characterization system, steam, water, nitrogen, and compressed air systems, chemical feed system, fire protection system, drains, sumps and the waste disposal systems, and heating, ventilating, and air conditioning systems.

  3. An overview of the BWR ECCS strainer blockage issues

    International Nuclear Information System (INIS)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-01-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

  4. Waste Isolation Pilot Plant simulated RH TRU waste experiments: Data and interpretation pilot

    International Nuclear Information System (INIS)

    Molecke, M.A.; Argueello, G.J.; Beraun, R.

    1993-04-01

    The simulated, i.e., nonradioactive remote-handled transuranic waste (RH TRU) experiments being conducted underground in the Waste Isolation Pilot Plant (WIPP) were emplaced in mid-1986 and have been in heated test operation since 9/23/86. These experiments involve the in situ, waste package performance testing of eight full-size, reference RH TRU containers emplaced in horizontal, unlined test holes in the rock salt ribs (walls) of WIPP Room T. All of the test containers have internal electrical heaters; four of the test emplacements were filled with bentonite and silica sand backfill materials. We designed test conditions to be ''near-reference'' with respect to anticipated thermal outputs of RH TRU canisters and their geometrical spacing or layout in WIPP repository rooms, with RH TRU waste reference conditions current as of the start date of this test program. We also conducted some thermal overtest evaluations. This paper provides a: detailed test overview; comprehensive data update for the first 5 years of test operations; summary of experiment observations; initial data interpretations; and, several status; experimental objectives -- how these tests support WIPP TRU waste acceptance, performance assessment studies, underground operations, and the overall WIPP mission; and, in situ performance evaluations of RH TRU waste package materials plus design details and options. We provide instrument data and results for in situ waste container and borehole temperatures, pressures exerted on test containers through the backfill materials, and vertical and horizontal borehole-closure measurements and rates. The effects of heat on borehole closure, fracturing, and near-field materials (metals, backfills, rock salt, and intruding brine) interactions were closely monitored and are summarized, as are assorted test observations. Predictive 3-dimensional thermal and structural modeling studies of borehole and room closures and temperature fields were also performed

  5. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  6. Summary of uranium refining and conversion pilot plant at Ningyo-toge works

    International Nuclear Information System (INIS)

    Iwata, Ichiro

    1981-01-01

    In the Ningyo-toge works, Power Reactor and Nuclear Fuel Development Corp., the construction of the uranium refining and conversion pilot plant was completed, and the operation will be started after the various tests based on the related laws. As for the uranium refining in Japan, the PNC process by wet refining method has been developed since 1958. The history of the development is described. It was decided to construct the refining and conversion pilot plant with 200 t uranium/year capacity as the comprehensive result of the development. This is the amount sufficient to supply UF 6 to the uranium enrichment pilot plant in Ningyo-toge. The building for the refining and conversion pilot plant is a three-story ferro-concrete building with the total floor area of about 13,000 m 2 . The raw materials are the uranium ore produced in Ningyo-toge and the yellow cakes from abroad. Uranyl sulfate solution is obtained by solvent extraction using an extraction tower or a mixer-settler. The following processes are electrolytic reduction, precipitation of uranium tetrafluoride, filtration, drying, dehydration and UF 6 conversion. The fluorine for UF 6 conversion is produced by the facility in the plant. The operation of the pilot plant will be started in the latter half of the fiscal year 1981, the batch operation is carrried out in 1982, and the continuous operation from 1983. (Kako, I.)

  7. The design of a continuous ion-exchange pilot plant for the recovery of uranium from partially clarified solutions

    International Nuclear Information System (INIS)

    Cloete, F.L.D.

    1980-01-01

    A preliminary design is given for a pilot plant to recover uranium from partially clarified slime pulp by continuous ion exchange. Process and plant-design methods are indicated briefly, and an outline is given of experimental work that should be undertaken before the start-up of the pilot plant

  8. General Atomic reprocessing pilot plant: description and results of initial testing

    International Nuclear Information System (INIS)

    1977-12-01

    In June 1976 General Atomic completed the construction of a reprocessing head-end cold pilot plant. In the year since then, each system within the head end has been used for experiments which have qualified the designs. This report describes the equipment in the plant and summarizes the results of the initial phase of reprocessing testing

  9. Vitrification of plutonium at Rocky Flats the argument for a pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Moore, L. [Rocky Mountain Peace Center, Boulder, CO (United States)

    1996-05-01

    Current plans for stabilizing and storing the plutonium at Rocky Flats Plant fail to put the material in a form suitable for disposition and resistant to proliferation. Vitrification should be considered as an alternate technology. The vitrification should begin with a small-scale pilot plant.

  10. Performance and Model Calibration of R-D-N Processes in Pilot Plant

    DEFF Research Database (Denmark)

    de la Sota, A.; Larrea, L.; Novak, L.

    1994-01-01

    This paper deals with the first part of an experimental programme in a pilot plant configured for advanced biological nutrient removal processes treating domestic wastewater of Bilbao. The IAWPRC Model No.1 was calibrated in order to optimize the design of the full-scale plant. In this first phas...

  11. Dismantling of an alpha contaminated hot cell at the Marcoule Pilot Plant

    International Nuclear Information System (INIS)

    Tachon, M.

    1988-01-01

    For the remodeling of Marcoule Pilot Plant, the cell 82: old unit for plutonium solution purification by extraction, was dismantled. About 42 tons of wastes were evacuated. Some wastes wen decontaminated by mechanical means other wastes with higher residual activity were stored for subsequent processing. The operation shows that dismantling of a hot cell is possible even if incorporated in an operating plant [fr

  12. Output-Feedback Model Predictive Control of a Pasteurization Pilot Plant based on an LPV model

    Science.gov (United States)

    Karimi Pour, Fatemeh; Ocampo-Martinez, Carlos; Puig, Vicenç

    2017-01-01

    This paper presents a model predictive control (MPC) of a pasteurization pilot plant based on an LPV model. Since not all the states are measured, an observer is also designed, which allows implementing an output-feedback MPC scheme. However, the model of the plant is not completely observable when augmented with the disturbance models. In order to solve this problem, the following strategies are used: (i) the whole system is decoupled into two subsystems, (ii) an inner state-feedback controller is implemented into the MPC control scheme. A real-time example based on the pasteurization pilot plant is simulated as a case study for testing the behavior of the approaches.

  13. Test operation of the uranium ore processing pilot plant and uranium conversion plant

    International Nuclear Information System (INIS)

    Suh, I.S.; Lee, K.I.; Whang, S.T.; Kang, Y.H.; Lee, C.W.; Chu, J.O.; Lee, I.H.; Park, S.C.

    1983-01-01

    For the guarantee of acid leaching process of the Uranium Ore Processing Pilot Plnat, the KAERI team performed the test operation in coorperation with the COGEMA engineers. The result of the operation was successful achieving the uranium leaching efficiency of 95%. Completing the guarentee test, a continuous test operation was shifted to reconform the reproducibility of the result and check the functions of every units of the pilot plant feeding the low-grade domestic ore, the consistency of the facility was conformed that the uranium can easily be dissolved out form the ore between the temperature range of 60degC-70degC for two hours of leaching with sulfuric acid and could be obtained the leaching efficiency of 92% to 95%. The uranium recovery efficiencies for the processes of extraction and stripping were reached to 99% and 99.6% respectively. As an alternative process for the separation of solid from the ore pulp, four of the Counter Current Decanters were shifted replacing the Belt Filter and those were connected in a series, which were not been tested during the guarantee operation. It was found out that the washing efficiencies of the ore pulp in each tests for the decanters were proportionally increased according to the quantities of the washing water. As a result of the test, it was obtained that washing efficiencies were 95%, 85%, 83% for the water to ore ratio of 3:1, 2:1, 1.5:1 respectively. (Author)

  14. Development of some operations in technological flowsheet for spent VVER fuel reprocessing at a pilot plant

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Galkin, B.Ya; Lyubtsev, R.I.; Romanovskii, V.N.; Velikhov, E.P.

    1981-01-01

    The fuel reprocessing pilot plants for high active materials would permit the study and development or particular processing steps and flowsheet variations; in some cases, these experimental installations realize on a small scale practically all technological chains of large reprocessing plants. Such a fuel reprocessing pilot plant with capacity of 3 kg U/d has been built at V. G. Khlopin Radium Institute. The pilot plant is installed in the hot cell of radiochemical compartment, and is composed of the equipments for fuel element cutting and dissolving, the preparation of feed solution (clarification, correction), extraction reprocessing and the production of uranium, plutonium and neptunium concentrates, the complex processing of liquid and solid wastes and a special unit for gas purification and analysis. In the last few years, a series of experiments have been carried out on the reprocessing of spent VVER fuel. (J.P.N.)

  15. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  16. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  17. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  18. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  19. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  20. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  1. Technical Description Lillgrund Wind Power Plant. Lillgrund Pilot Project

    Energy Technology Data Exchange (ETDEWEB)

    Jeppsson, Joakim; Larsen, Poul Erik; Larsson, Aake [Vattenfall Vindkraft AB, Stockholm (Sweden)

    2008-09-15

    Lillgrund offshore wind power plant comprises 48 wind turbines, each rated at 2.3 MW, bringing the total wind farm capacity to 110 MW. The Lillgrund offshore wind power plant is located in a shallow area of Oeresund, 7 km off the coast of Sweden and 7 km south from the Oeresund Bridge connecting Sweden and Denmark. An average wind speed of around 8,5 m/s at hub height, combined with a relatively low water depth of 4 to 8 meters makes it economically feasible to build here. Vattenfall Vindkraft AB is the owner and operator of Lillgrund offshore wind power plant. Lillgrund is a Swedish pilot project supported by the Swedish Energy Agency. The bidding process was completed during 2005 and the offshore power plant was constructed in the period 2006 to 2007. The wind farm was constructed on time and has now been successfully operational since December 2007. There is, however, always potential for improvement and the aim of this report has been to determine and highlight these areas. It is worth noting out that only the electrical system and the foundations are tailor made at offshore wind power plants. The wind turbines are more or less standard products with none or very limited possibilities for project specific design changes. Geotechnical investigations are expensive and it can be difficult to balance the risks as well as the benefits of this expense in the early phases of a large infrastructure project. As a whole, the geotechnical surveys at Lillgrund proved to be useful. They identified potential issues, such as the fact that extra excavation was required for two of the foundations. It also revealed the location of a small number of boulders that would have to be removed. Vattenfall requested a complete study of the electrical system for Lillgrund to be delivered with the bids. That request was not met. Instead Siemens Wind Power began a complete electrical system study after being awarded the Contract. The electrical system study was completed during the

  2. Remote maintenance demonstration tests at a pilot plant for high level waste vitrification

    International Nuclear Information System (INIS)

    Selig, M.

    1984-01-01

    The remote maintenance and replacement technique designed for a radioactive vitrification plant have been developed and tested in a full scale handling mockup and in an inactive pilot plants by the Central Engineering Department of the Karlsruhe Nuclear Research Center. As a result of the development work and the tests it has been proved that the remote maintenance technique and remote handling equipment can be used without any technical problems and are suited for application in a radioactive waste vitrification plant

  3. Raft River binary-cycle geothermal pilot power plant final report

    Energy Technology Data Exchange (ETDEWEB)

    Bliem, C.J.; Walrath, L.F.

    1983-04-01

    The design and performance of a 5-MW(e) binary-cycle pilot power plant that used a moderate-temperature hydrothermal resource, with isobutane as a working fluid, are examined. Operating problems experienced and solutions found are discussed and recommendations are made for improvements to future power plant designs. The plant and individual systems are analyzed for design specification versus actual performance figures.

  4. Operation result of 40kW class MCFC pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Saitoh, H.; Hatori, S.; Hosaka, M.; Uematsu, H. [Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo (Japan)

    1996-12-31

    Ishikawajima-Harima Heavy Industries Co., Ltd. developed unique Molten Carbonate Fuel Cell (MCFC) system based on our original concept. To demonstrate the possibility of this system, based on MCFC technology of consigned research from New Energy and Industrial Technology Development Organization (NEDO) in Japan, we designed 40kW class MCFC pilot plant which had all equipments required as a power plant and constructed in our TO-2 Technical Center. This paper presents the test results of the plant.

  5. 10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-25

    Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)

  6. Strategic pilot for operator support system in nuclear power plant - design considerations

    International Nuclear Information System (INIS)

    Bucur, I.; Tatar, F.

    1999-01-01

    In order to improve the plant operational safety the development of an Operator Support System (OSS) is required. This system is intended to process data from nuclear systems and to provide adequate outputs to the plant operation staff. Before implementing this system, a strategic pilot should be produced as a demonstration of the technology. The strategic pilot could be considered as a means of building both skills and credibility in development and implementation of OSS. In any organization this project should be under plant management control with operation group involvement. This paper describes the managerial tasks that should be carried out to define, build and implement such a module. The main objectives, the functional requirements and the benefits of pilot implementation are revealed. Furthermore, the problem relating to the background at CNE-PROD Cernavoda is analyzed and the present achievements are pointed out. (authors)

  7. Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

    Energy Technology Data Exchange (ETDEWEB)

    Paar, H; Narodoslawsky, M; Moser, A [Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie

    1990-10-10

    Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

  8. Development of the Risk-Based Inspection Techniques and Pilot Plant Activities

    International Nuclear Information System (INIS)

    Phillips, J.H.

    1997-01-01

    Risk-based techniques have been developed for commercial nuclear power plants. System boundaries and success criteria is defined using the probabilistic risk analysis or probabilistic safety analysis developed to meet the individual plant evaluation. Final ranking of components is by a plant expert panel similar to the one developed for maintenance rule. Components are identified as being high risk-significant or low-risk significant. Maintenance and resources are focused on those components that have the highest risk-significance. The techniques have been developed and applied at a number of pilot plants. Results from the first risk-based inspection pilot plant indicates that safety due to pipe failure can be doubled while the inspection reduced to about 80% when compared with current inspection programs. The reduction in inspection reduces the person-rem exposure resulting in further increases in safety. These techniques have been documented in publication by the ASME CRTD

  9. Verification of criticality Safety for ETRR-2 Fuel Manufacturing pilot Plant (FMPP) at Inshas

    International Nuclear Information System (INIS)

    Aziz, M.; Gadalla, A.A.; Orabi, G.

    2006-01-01

    The criticality safety of the fuel manufacturing pilot plant (FMPP) at inshas is studied and analyzed during normal and abnormal operation conditions. the multiplication factor during all stages of the manufacturing processes is determined. several accident scenarios were simulated and the criticality of these accidents were investigated. two codes are used in the analysis : MCNP 4 B code, based on monte Carlo method, and CITATION code , based on diffusion theory. the results are compared with the designer calculations and satisfactory agreement were found. the results of the study indicated that the safety of the fuel manufacturing pilot plant is confirmed

  10. Macroscopic mass and energy balance of a pilot plant anaerobic bioreactor operated under thermophilic conditions.

    Science.gov (United States)

    Espinosa-Solares, Teodoro; Bombardiere, John; Chatfield, Mark; Domaschko, Max; Easter, Michael; Stafford, David A; Castillo-Angeles, Saul; Castellanos-Hernandez, Nehemias

    2006-01-01

    Intensive poultry production generates over 100,000 t of litter annually in West Virginia and 9 x 10(6) t nationwide. Current available technological alternatives based on thermophilic anaerobic digestion for residuals treatment are diverse. A modification of the typical continuous stirred tank reactor is a promising process being relatively stable and owing to its capability to manage considerable amounts of residuals at low operational cost. A 40-m3 pilot plant digester was used for performance evaluation considering energy input and methane production. Results suggest some changes to the pilot plant configuration are necessary to reduce power consumption although maximizing biodigester performance.

  11. Integrated automation system for a pilot plant for energy conversion using PEMFCs

    International Nuclear Information System (INIS)

    Culcer, Mihai; Iliescu, Mariana; Raceanu, Mircea; Stanciu, Vasile; Stefanescu, Ioan; Enache, Adrian; Lazaro, Pavel Gabriel; Lazaroiu, Gheorghe; Badea, Adrian

    2007-01-01

    Based on Hydrogen and Fuel Cells researches and technological capabilities achieved in the National R and D Programs, ICIT Rm. Valcea built an experimental-demonstrative pilot plant for energy conversion using hydrogen PEMFCs. This pilot plant consists of a fuel processor based on steam methane reforming (SMR) process, a hydrogen purification unit, a PEM fuel cells stack (FCS) and a power electronics unit. The paper deals with the dedicated controlling system that provides automated data acquisition, manual or on-line operational control, gas management, humidification, temperature and flow controls. (authors)

  12. Development of a computer systems for operational data acquisition of uranium isotopic enrichment pilot plant

    International Nuclear Information System (INIS)

    Maia, W.M.C.

    1985-01-01

    A pilot plant for uranium enrichment using the jet nozzle process was transfered from Federal Republic of Germany to Brazil, to train Brazilian technicist in its operation and to improve the process. This pilot plant is monitored by a data acquisition system and the possibility of faulty events would cause serious dificulties, as far as maintenance is concerned (for instance, unvailable special components). It is described the development of a new system, which is proposed in order to minimize difficulties with maintenance that utilizes in the assembling integrated circuits of large scale of integration. It is controlled by a microcomputer. (Author) [pt

  13. Waste Isolation Pilot Plant transuranic wastes experimental characterization program: executive summary

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1978-11-01

    A general overview of the Waste Isolation Pilot Plant transuranic wastes experimental characterization program is presented. Objectives and outstanding concerns of this program are discussed. Characteristics of transuranic wastes are also described. Concerns for the terminal isolation of such wastes in a deep bedded salt facility are divided into two phases, those during the short-term operational phase of the facility, and those potentially occurring in the long-term, after decommissioning of the repository. An inclusive summary covering individual studies, their importance to the Waste Isolation Pilot Plant, investigators, general milestones, and comments are presented

  14. Advanced chemistry management system to optimize BWR chemistry control

    International Nuclear Information System (INIS)

    Maeda, K.; Nagasawa, K.

    2002-01-01

    BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

  15. Continued oversight of the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Peake, R. Thomas

    2014-01-01

    The United States Environmental Protection Agency (EPA) developed environmental standards applicable to the disposal of defence-related transuranic wastes at the US Department of Energy's (DOE) Waste Isolation Pilot Plant (WIPP). By statute, EPA also serves as the regulator and implements these standards at WIPP, which has been in operation since 1999. The general environmental standards are set forth in the Agency's 40 Code of Federal Regulations (CFR), Part 191 Environmental Radiation Protection Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes (US NARA, 1985). These standards are implemented by site-specific compliance criteria at 40 CFR 194 (US NARA, 1996). The repository waste area is ∼650 meters below ground surface in a thick bedded salt formation that dips from west to east at ∼1 deg.. WIPP is located in the Chihuahuan Desert of south-eastern New Mexico, where the annual precipitation averages between 25 and 40 centimetres and there is high evapotranspiration. Much of the area around WIPP is federal land, managed by the Bureau of Land Management, and the area is sparsely populated. The transuranic waste disposed of at WIPP consists of materials such as radioactive sludges, soils and laboratory materials (e.g. chemical mixtures, contaminated glove boxes, paper and glass). Wastes are typically not treated unless necessary for shipping purposes (e.g. to limit hydrogen build-up). The waste is contaminated with plutonium, americium and other radionuclides, including some caesium and strontium. Transuranic waste is defined as waste with radionuclides heavier than uranium containing more than 3 700 Bq (100 nanocuries) of alpha-emitting transuranic isotopes per gram of waste; isotopes must have half-lives greater than 20 years. The WIPP Land Withdrawal Act limits the total disposal volume to ∼177 000 cubic meters (6.2 million cubic feet) and creates two categories of waste based on operational

  16. Laboratory Testing of Waste Isolation Pilot Plant Surrogate Waste Materials

    Science.gov (United States)

    Broome, S.; Bronowski, D.; Pfeifle, T.; Herrick, C. G.

    2011-12-01

    The Waste Isolation Pilot Plant (WIPP) is a U.S. Department of Energy geological repository for the permanent disposal of defense-related transuranic (TRU) waste. The waste is emplaced in rooms excavated in the bedded Salado salt formation at a depth of 655 m below the ground surface. After emplacement of the waste, the repository will be sealed and decommissioned. WIPP Performance Assessment modeling of the underground material response requires a full and accurate understanding of coupled mechanical, hydrological, and geochemical processes and how they evolve with time. This study was part of a broader test program focused on room closure, specifically the compaction behavior of waste and the constitutive relations to model this behavior. The goal of this study was to develop an improved waste constitutive model. The model parameters are developed based on a well designed set of test data. The constitutive model will then be used to realistically model evolution of the underground and to better understand the impacts on repository performance. The present study results are focused on laboratory testing of surrogate waste materials. The surrogate wastes correspond to a conservative estimate of the degraded containers and TRU waste materials after the 10,000 year regulatory period. Testing consists of hydrostatic, uniaxial, and triaxial tests performed on surrogate waste recipes that were previously developed by Hansen et al. (1997). These recipes can be divided into materials that simulate 50% and 100% degraded waste by weight. The percent degradation indicates the anticipated amount of iron corrosion, as well as the decomposition of cellulosics, plastics, and rubbers. Axial, lateral, and volumetric strain and axial and lateral stress measurements were made. Two unique testing techniques were developed during the course of the experimental program. The first involves the use of dilatometry to measure sample volumetric strain under a hydrostatic condition. Bulk

  17. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    Salaices, M.; Salaices, E.; Ovando, R.; Esquivias, J.

    2011-11-01

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  18. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  19. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  20. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  1. Investigation of Parameters Affecting Gypsum Dewatering Properties in a Wet Flue Gas Desulphurization Pilot Plant

    DEFF Research Database (Denmark)

    Hansen, Brian Brun; Kiil, Søren

    2012-01-01

    of impurities (0.002 M Al2F6; 50 g quartz/L; 0.02 M Al3+, and 0.040 M Mg2+) were investigated. In addition, slurry from a full-scale wet FGD plant, experiencing formation of flat shaped crystals and poor gypsum dewatering properties, was transferred to the pilot plant to test if the plant would now start...... to time. In this work, the particle size distribution, morphology, and filtration rate of wet FGD gypsum formed in a pilot-scale experimental setup, operated in forced oxidation mode, have been studied. The influence of holding tank residence time (10–408 h), solids content (30–169 g/L), and the presence...... to produce low quality gypsum. The crystals formed in the pilot plant, on the basis of the full-scale slurry did, however, show acceptable filtration rates and crystal morphologies closer to the prismatic crystals from after pilot plant experiments with demineralized water. The gypsum slurry filtration rates...

  2. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    International Nuclear Information System (INIS)

    Menard, J.E.; Bromberg, L.; Brown, T.; Burgess, Thomas W.; Dix, D.; Gerrity, T.; Goldston, R.J.; Hawryluk, R.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G.H.; Neumeyer, C.L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.G.; Zarnstorff, M.C.

    2011-01-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  3. Water-gas shift (WGS) Operation of Pre-combustion CO2 Capture Pilot Plant at the Buggenum IGCC

    NARCIS (Netherlands)

    Van Dijk, H.A.J.; Damen, K.; Makkee, M.; Trapp, C.

    2014-01-01

    In the Nuon/Vattenfall CO2 Catch-up project, a pre-combustion CO2 capture pilot plant was built and operated at the Buggenum IGCC power plant, the Netherlands. The pilot consist of sweet water-gas shift, physical CO2 absorption and CO2 compression. The technology performance was verified and

  4. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  5. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  6. Arsenic pilot plant operation and results:Weatherford, Oklahoma.

    Energy Technology Data Exchange (ETDEWEB)

    Aragon, Malynda Jo; Arora, H. (Narasimhan Consulting Services Inc., Phoenix, Arizona); Karori, Saqib (Narasimhan Consulting Services Inc., Phoenix, Arizona); Pathan, Sakib (Narasimhan Consulting Services Inc., Phoenix, Arizona)

    2007-05-01

    Narasimhan Consulting Services, Inc. (NCS), under a contract with the Sandia National Laboratories (SNL), designed and operated pilot scale evaluations of the adsorption and coagulation/filtration treatment technologies aimed at meeting the recently revised arsenic maximum contaminant level (MCL) for drinking water. The standard of 10 {micro}g/L (10 ppb) is effective as of January 2006. The pilot demonstration is a project of the Arsenic Water Technology Partnership program, a partnership between the American Water Works Association Research Foundation (AwwaRF), SNL and WERC (A Consortium for Environmental Education and Technology Development). The pilot evaluation was conducted at Well 30 of the City of Weatherford, OK, which supplies drinking water to a population of more than 10,400. Well water contained arsenic in the range of 16 to 29 ppb during the study. Four commercially available adsorption media were evaluated side by side for a period of three months. Both adsorption and coagulation/filtration effectively reduced arsenic from Well No.30. A preliminary economic analysis indicated that adsorption using an iron oxide media was more cost effective than the coagulation/ filtration technology.

  7. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  8. Pilot-scale Biogas Plant for the Research and Development of New Technologies

    Directory of Open Access Journals (Sweden)

    Ivan Simeonov

    2012-09-01

    Full Text Available Тhe paper describes a new pilot-scale biogas plant of the Institute of Microbiology - Bulgarian Academy of Sciences. The equipment includes: a 100 L pilot bioreactor, a 200 L metal gasholder, sensors, actuators, a two-level automatic process monitoring and control system, a fire and explosion protection system and two web cameras. The monitoring and control system is composed on the lower level of a controller Beckhoff, and on the higher level - of a PC with specialized software (under development. The pilot biogas plant is designed to work out and scale up various anaerobic digestion (AD technologies based on different types of feedstock. All the data will be stored on the PC for quick reference and possibly data mining, parameter identification and verification of different AD mathematical models.

  9. Radioactive Waste Treatment and Conditioning Using Plasma Technology Pilot Plant: Testing and Commissioning

    International Nuclear Information System (INIS)

    Rafizi Salihuddin; Rohyiza Baan; Norasalwa Zakaria

    2016-01-01

    Plasma pilot plant was commissioned for research and development program on radioactive waste treatment. The plant is equipped with a 50 kW direct current of non-transferred arc plasma torch which mounted vertically on top of the combustion chamber. The plant also consists of a dual function chamber, a water cooling system, a compress air supply system and a control system. This paper devoted the outcome after testing and commissioning of the plant. The problems arise was discussed in order to find the possible suggestion to overcome the issues. (author)

  10. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  11. Pilot plant studies of the bioconversion of cellulose and production of ethanol

    Energy Technology Data Exchange (ETDEWEB)

    Wilke, C.R.

    1977-09-30

    Work for the period July 1 to September 30, 1977 is summarized briefly. Results of the following studies are reported: analysis and evaluation of potential raw materials--chemical analysis of the Kudzu plant and effect of NO/sub x/ pretreatments on the hydrolysis of wheat straw; utilization of hemicellulose sugars; process design and economic studies--hydrolysis process and ethanol fermentation; pilot plant process development and design studies--enhanced cellulase production and continuous hydrolysis. (JGB)

  12. The design of a modular pilot plant based on the adsorber loop concept

    International Nuclear Information System (INIS)

    Koske, P.H.; Ohlrogge, K.

    1984-01-01

    The main design criteria for a pilot plant producing about 100 t uranium per year from seawater are discussed. The application of the ''adsorber loop concept'' for the contact between seawater and the adsorber granulate enables the employment of high seawater velocities. The seawater flow is accomplished by active pumping and the plant is supposed to be operating far from shores. Besides some informations on the theoretical background the essential engineering considerations are presented. (orig.) [de

  13. Modernization and enlarging of the Marcoule pilot plant for R and D

    International Nuclear Information System (INIS)

    Calame-Longjean, A.; Revol, G.; Roux, J.P.; Ranger, G.

    1987-01-01

    The aim of the pilot plant is the testing of process and equipment in actual conditions with spent fuels on a half-industrial scale and for a significant time. From 1963 to 1983 more than 11t of spent fuels (mainly from fast reactors) were reprocessed. Since 1983 is modernized and enlarged and the new plant of the TOR project (treatment of oxides from fast reactors) are described [fr

  14. Use of phosphorus release batch tests for modelling an EBPR pilot plant

    DEFF Research Database (Denmark)

    Tykesson, E.; Aspegren, H.; Henze, Mogens

    2002-01-01

    The aim of this study was to evaluate how routinely performed phosphorus release tests could be used when modelling enhanced biological phosphorus removal (EBPR) using activated sludge models such as ASM2d. A pilot plant with an extensive analysis programme was used as basis for the simulations...

  15. Drying and purification of natural gas by clinoptilolite on an experimental pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Tsitsishvili, G V; Urotadze, S L; Lukin, V D; Bagirov, R M

    1976-02-01

    The paper deals with the process of the drying and purification of natural gas from CO/sub 2/ on an experimental pilot plant using the natural zeolite clinoptilolite. On the basis of the obtained data the dynamic activity of clinoptilolite against water and CO/sub 2/ has been calculated.

  16. The Waste Isolation Pilot Plant: a potential solution for the disposal of transuranic waste

    National Research Council Canada - National Science Library

    National Research Council Staff; Commission on Geosciences, Environment and Resources; Division on Earth and Life Studies; National Research Council; National Academy of Sciences

    ... Isolation Pilot Plant Board on Radioactive Waste Management Commission on Geosciences, Environment, and Resources National Research Council NATIONAL ACADEMY PRESS Washington, D.C. 1996 i Copyrighttrue Please breaks inserted. are Page files. accidentally typesetting been have may original from the errors not typographic original retained, and from the c...

  17. Hydraulic testing of Salado Formation evaporites at the Waste Isolation Pilot Plant site: Second interpretive report

    Energy Technology Data Exchange (ETDEWEB)

    Beauheim, R.L. [Sandia National Labs., Albuquerque, NM (United States); Roberts, R.M.; Dale, T.F.; Fort, M.D.; Stensrud, W.A. [INTERA, Inc., Austin, TX (United States)

    1993-12-01

    Pressure-pulse, constant-pressure flow, and pressure-buildup tests have been performed in bedded evaporites of the Salado Formation at the Waste Isolation Pilot Plant (WIPP) site to evaluate the hydraulic properties controlling brine flow through the Salado. Transmissivities have been interpreted from six sequences of tests conducted on five stratigraphic intervals within 15 m of the WIPP underground excavations.

  18. Continuous operation of a pilot plant for the production of beryllium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Costa, T C; Amaral, S; Silveira, C M.S.; de Oliveira, A P [Instituto de Tecnologia, Governador Valadares (Brazil)

    1975-12-01

    A method of obtaining beryllium oxide with a purity of 99,2% was developed in a pilot plant with a capacity of 7 tons per month destined to operate continuously. The operation market prospects and control of production with the objective of obtaining internacional technical grade beryllium oxide are discussed.

  19. Retrieval of canistered experimental waste at the waste isolation pilot plant

    International Nuclear Information System (INIS)

    Stinebaugh, R.E.

    1979-07-01

    To assess the suitability of bedded salt for nuclear waste disposal, an extensive experimental program will be implemented at the Waste Isolation Pilot Plant. In order to evaluate experimental results, it will be necessary to recover certain of these experiments for postmortem examination and analysis. This document describes the equipment and procedures used to effect recovery of one category of WIPP experiments

  20. Waste Isolation Pilot Plant Geotechnical Analysis Report for July 2005 - June 2006, Volume 2, Supporting Data

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2007-03-25

    This report is a compilation of geotechnical data presented as plots for each active instrument installed in the underground at the Waste Isolation Pilot Plant (WIPP) through June 30, 2006. A summary of the geotechnical analyses that were performed using the enclosed data is provided in Volume 1 of the Geotechnical Analysis Report (GAR).

  1. Continuous operation of a pilot plant for the production of beryllium oxide

    International Nuclear Information System (INIS)

    Costa, T.C.; Amaral, S.; Silveira, C.M.S.; Oliveira, A.P. de

    1975-01-01

    A method of obtaining beryllium oxide with a purity of 99,2% was developed in a pilot plant with a capacity of 7 tons per month destined to operate continuously. The operation market prospects and control of production with the objective of obtaining internacional technical grade beryllium oxide are discussed [pt

  2. Electrodialytic remediation of CCA-treated waste wood in a 2 m3 pilot plant

    DEFF Research Database (Denmark)

    Christensen, Iben Vernegren; Pedersen, Anne Juul; Ottosen, Lisbeth M.

    2006-01-01

    Waste wood that has been treated with chromated-copper-arsenate (CCA) poses a potential environmental problem due to the content of copper, chromium and arsenic. A pilot plant for electrodialytic remediation of up to 2 m3 wood has been designed and tested and the results are presented here. Sever...

  3. WIPP conceptual design report. Addendum A. Design calculations for Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    1977-04-01

    The design calculations for the Waste Isolation Pilot Plant (WIPP) are presented. The following categories are discussed: general nuclear calculations; radwaste calculations; structural calculations; mechanical calculations; civil calculations; electrical calculations; TRU waste surface facility time and motion analysis; shaft sinking procedures; hoist time and motion studies; mining system analysis; mine ventilation calculations; mine structural analysis; and miscellaneous underground calculations

  4. Radioactive Waste Disposal Pilot Plant concept for a New Mexico site

    International Nuclear Information System (INIS)

    Weart, W.D.

    1976-01-01

    Twenty years of investigation have shown that disposal of nuclear wastes in deep salt formations is the surest means of isolating these wastes from the biosphere for the extremely long period of time required. A large scale demonstration of this capability will soon be provided by a Radioactive Waste Disposal Pilot Plant (RWDPP) to be developed in southeastern New Mexico. Initially, the pilot plant will accept only ERDA generated waste; high level waste from the commercial power reactor fuel cycle will eventually be accommodated in the pilot plant and the initial RWDPP design will be compatible with this waste form. Selection of a specific site and salt horizon will be completed in June 1976. Conceptual design of the RWDPP and assessment of its environmental impact will be completed by June 1977. Construction is expected to start in 1978 with first waste accepted in 1982. The present concept develops disposal areas for all nuclear waste types in a single salt horizon about 800 meters deep. This single level can accommodate all low level and high level waste generated in the United States through the year 2010. A major constraint on the RWDPP design is the ERDA requirement that all waste be ''readily'' retrievable during the duration of pilot plant operation

  5. Not-from-concentrate pilot plant ‘Wonderful’ cultivar pomegranate juice changes: Volatiles

    Science.gov (United States)

    Pilot plant ultrafiltration was used to mimic the dominant U.S. commercial pomegranate juice extraction method (hydraulic pressing whole fruit), to deliver a not-from-concentrate (NFC) juice that was high-temperature short-time pasteurized and stored at 4 and 25 °C. Recovered were 46 compounds, of ...

  6. Waste Isolation Pilot Plant Geotechnical Analysis Report for July 2005 - June 2006, Volume 2, Supporting Data

    International Nuclear Information System (INIS)

    2007-01-01

    This report is a compilation of geotechnical data presented as plots for each active instrument installed in the underground at the Waste Isolation Pilot Plant (WIPP) through June 30, 2006. A summary of the geotechnical analyses that were performed using the enclosed data is provided in Volume 1 of the Geotechnical Analysis Report (GAR).

  7. Hydraulic testing of Salado Formation evaporites at the Waste Isolation Pilot Plant site: Second interpretive report

    International Nuclear Information System (INIS)

    Beauheim, R.L.; Roberts, R.M.; Dale, T.F.; Fort, M.D.; Stensrud, W.A.

    1993-12-01

    Pressure-pulse, constant-pressure flow, and pressure-buildup tests have been performed in bedded evaporites of the Salado Formation at the Waste Isolation Pilot Plant (WIPP) site to evaluate the hydraulic properties controlling brine flow through the Salado. Transmissivities have been interpreted from six sequences of tests conducted on five stratigraphic intervals within 15 m of the WIPP underground excavations

  8. Pasteurization of strawberry puree using a pilot plant pulsed electric fields (PEF) system

    Science.gov (United States)

    The processing of strawberry puree by pulsed electric fields (PEF) in a pilot plant system has never been evaluated. In addition, a method does not exist to validate the exact number and shape of the pulses applied during PEF processing. Both buffered peptone water (BPW) and fresh strawberry puree (...

  9. Resource conservation and recovery act draft hazardous waste facility permit: Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    1993-08-01

    Volume I contains the following attachments for Module II: waste analysis plan; quality assurance program plan for the Waste Isolation Pilot Plant (WIPP) Experiment Waste Characterization Program(QAPP); WIPP Characterization Sampling and Analysis Guidance Manual (Plan)(SAP); and no migration Determination Requirement Summary (NMD)

  10. Developments in the pre-combustion CO2 capture pilot plant at the Buggenum IGCC

    NARCIS (Netherlands)

    Damen, K.; Gnutek, R.; Kaptein, J.; Nannan, N.R.; Oyarzun, B.; Trapp, C.; Colonna, P.; Van Dijk, E.; Gross, J.; Bardow, A.

    2011-01-01

    N.V. Nuon (part of the Vattenfall Group) operates an IGCC in Buggenum and is developing a multi-fuel IGCC with CO2 capture and storage (Nuon Magnum) in Eemshaven, the Netherlands. In order to prepare for large-scale application of CO2 capture and storage, a CO2 capture pilot plant is constructed at

  11. Improvement of water treatment pilot plant with Moringa oleifera extract as flocculant agent.

    Science.gov (United States)

    Beltrán-Heredia, J; Sánchez-Martín, J

    2009-05-01

    Moringa oleifera extract is a high-capacity flocculant agent for turbidity removal in surface water treatment. A complete study of a pilot-plant installation has been carried out. Because of flocculent sedimentability of treated water, a residual turbidity occured in the pilot plant (around 30 NTU), which could not be reduced just by a coagulation-flocculation-sedimentation process. Because of this limitation, the pilot plant (excluded filtration) achieved a turbidity removal up to 70%. A slow sand filter was put in as a complement to installation. A clogging process was characterized, according to Carman-Kozeny's hydraulic hypothesis. Kozeny's k parameter was found to be 4.18. Through fouling stages, this k parameter was found to be up to 6.36. The obtained data are relevant for the design of a real filter in a continuous-feeding pilot plant. Slow sand filtration is highly recommended owing to its low cost, easy-handling and low maintenance, so it is a very good complement to Moringa water treatment in developing countries.

  12. Double stage dry-wet-fermentation - start-up of a pilot biogas plant

    International Nuclear Information System (INIS)

    Buschmann, Jeannette; Busch, Gunter; Burkhardt, Marko

    2009-01-01

    The Brandenburg University of Technology (BTU) has developed a double stage dry-wet fermentation process for fast and safe anaerobic degradation. Originally designed for treatment of organic wastes, this process allows using a wide variety of solid biodegradable materials. The dividing of hydrolysis and methanation in this process, allows an optimization of the different steps of biogas generation separately. The main advantages of the process are the optimum process control, an extremely stable process operation and a high gas productivity and quality. Compared to conventional processes, the retention times within the percolation stage (hydrolysis) are reduced considerably. In cooperation with the engineering and consulting company GICON, the technology was qualified further to an industrial scale. In 2007 a pilot plant, and, simultaneously, an industrial plant were built by GICON based on this double stage technology. Based on practical experience from the operation of laboratory fermentation plants, the commissioning of the pilot plant was planned, controlled and monitored by our institution. The start-up of a biogas plant of this type focuses mainly on the inoculation the of methane reactor. The growth of microbial populations and generation of a stable biocenosis within the methane reactor is essential and affects the duration of starting period as well as the methanation efficiency a long time afterwards. This paper concerns with start-up of a pilot biogas plant and discusses particular occurrences and effects during this period. (author)

  13. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  14. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor; Analisis del accidente de la planta nucleoelectrica de Fukushima Daiichi en un reactor tipo ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Escorcia O, D. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Salazar S, E., E-mail: daniel.escorcia.ortiz@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2016-09-15

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  15. Planning of occupational dose reduction at BWR power plant by past dose record analysis combined with on-site workers' idea analysis

    International Nuclear Information System (INIS)

    Konno, T.; Taira, J.; Hayashida, T.; Suzuki, A.; Hayashi, K.; Kato, S.; Ishikawa, T.; Konno, T.; Hayashi, K.

    2011-01-01

    In order to establish a plan for occupational dose reduction at operating plants, outage inspection works that involve high-dose exposure were selected and a determination of the major causes of high-dose exposure made by plant-by-plant comparison of doses received during inspection works. The comparison was made to investigate the relationship between exposure and the volume of objects to be inspected, working time and man-hour of each work process and ambient dose rates at work areas. In parallel with this, an analysis has also been carried out on 400 data items in a questionnaire survey conducted on relevant individuals, including foremen, radiation safety personnel, on-site workers and plant designers regarding ideas for dose reduction methods. With combination of these two analyses, matters that require improvement will be highlighted, then modification of equipment or revision of work procedures necessary for occupational dose reduction will be planned by plant designers through review. (authors)

  16. Pilot plant UF6 to UF4 test operations report

    International Nuclear Information System (INIS)

    Bicha, W.J.; Fallings, M.; Gilbert, D.D.; Koch, G.E.; Levine, P.J.; McLaughlin, D.F.; Nuhfer, K.R.; Reese, J.C.

    1987-02-01

    The FMPC site includes a plant designed for the reduction of uranium hexafluoride (UF 6 ) to uranium tetrafluoride (UF 4 ). Limited operation of the upgraded reduction facility began in August 1984 and continued through January 19, 1986. A reaction vessel ruptured on that date causing the plant operation to be shut down. The DOE conducted a Class B investigation with the findings of the investigation board issued in preliminary form in May 1986 and as a final recommendation in July 1986. A two-phase restart of the plant was planned and implemented. Phase I included implementing safety system modifications, changing reaction vessel temperature control strategy, and operating the reduction plant under an 8-week controlled test. The results of the test period are the subject of this report. 41 figs., 11 tabs

  17. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  18. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  19. BWR steel containment corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  20. Safety study on nuclear heat utilization system - accident delineation and assessment on nuclear steelmaking pilot plant

    International Nuclear Information System (INIS)

    Yoshida, T.; Mizuno, M.; Tsuruoka, K.

    1982-01-01

    This paper presents accident delineation and assessment on a nuclear steelmaking pilot plant as an example of nuclear heat utilization systems. The reactor thermal energy from VHTR is transported to externally located chemical process plant employing helium-heated steam reformer by an intermediate heat transport loop. This paper on the nuclear steelmaking pilot plant will describe (1) system transients under accident conditions, (2) impact of explosion and fire on the nuclear reactor and the public and (3) radiation exposure on the public. The results presented in this paper will contribute considerably to understanding safety features of nuclear heat utilization system that employs the intermediate heat transport loop and the helium-heated steam reformer

  1. Experimental investigation of the chemical looping method on a 1 MW pilot plant

    International Nuclear Information System (INIS)

    Orth, Matthias

    2014-01-01

    Attempting to counteract the consequences of climate change, leading industrial nations have agreed on reducing their CO 2 emissions significantly. To reach these reduction goals, it is essential to reduce the CO 2 emissions in the field of energy conversion. This PHD thesis covers the field of chemical looping combustion, a technology that uses fossil fuels for energy conversion with inherent capture of CO 2 . Since the research regarding chemical looping had so far focused mainly on lab scale or small scale experiments, a 1 MW pilot plant has been erected at Technische Universitaet Darmstadt in order to investigate the process in a semi-industrial scale and to check the process efficiency with commercially usable equipment. This pilot consists of two interconnected fluidized bed reactors and has an overall height of more than 11 m. In this thesis, some experiments with ilmenite - used as the oxygen carrier - are explained. Furthermore, the design, erection and commissioning of the pilot plant are presented as well as the results of the first test campaigns. The evaluation of the latter proves that the process can be handled in the design configuration and that CO 2 can be safely captured in a pilot plant of this scale.

  2. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  3. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  4. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

  5. Pilot project concerning the establishment of a collective biomass conversion plant on the island of Mors

    International Nuclear Information System (INIS)

    1993-06-01

    This pilot project comprises a feasibility study in connection with plans to establish a biomass conversion plant, on the Danish island of Mors, which would provide methane to be used as fuel, in combination with natural gas, for a cogeneration plant serving six villages. The subjects of location, organization, the transportation of biomass, the design of the biomass conversion plant, economical aspects and conditions of the use of the methane are discussed as a basis for decisions in this respect. Environmental considerations are also dealt with. (AB)

  6. Diffusion bonded matrix of HGMF applied for BWR condensate water purification

    International Nuclear Information System (INIS)

    Soda, Fumitaka; Yukawa, Takao; Ito, Kazuyuki.

    1984-01-01

    High Gradient Magnetic Filter (HGMF) applied to the purification of power plant primary water has recently attracted much attention. In the application of HGMF to the water treatment of power plants, especially nuclear power plants, reliabillties of matrix (filtering medium) as well as removal performance for cruds (insoluble corrosion products) are considered to be important factors. To satisfy these factors, a new filtering medium named Diffision Bonded Matrix (DBM) has been developed and the test results are reported. Filtering efficiency and mechanical stiffness of DBM were examined using HGMF pilot test units consisting of 160 mm diameters x 240 mm length filter. The filtering velocity and the magnetic flux density used in this test were 800 m/h 5 kG, respectively. The filtering efficiencies and of 85-100% were obtained for artificial cruds for DBM. The DBM indicated slightly better filtering efficiency than for conventional wool matrix under the same filtering and matrix conditions. The DBM kept its original mechanical properties and very few pieces of fibers were broken off while the conventional wool matrix lost its volume elasticities and the considerable amount of fibers was broken off during the test operation. The results described here demonstrated the applicability of DBM for treatment of BWR primary water by High Gradient Magnetic Filter. (author)

  7. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  8. Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

    2011-07-01

    Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

  9. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  10. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    In 1985 an incident at Toledo Edison's Davis Besse plant caused the U.S. Nuclear Regulatory Commission (NRC) to re-evaluate the technical information that the utilities had readily available to support the design of their plants. The Design Basis programs, currently on going in most U.S. utilities, have been the nuclear industry's response to the needs identified by this re-evaluation. In order to understand the Design Basis programs which have been implemented by the U.S. nuclear utilities, it is necessary to understand the problem as it was perceived by the nuclear industry (the utilities, the original NSSS designers and the regulators) after the Davis-Besse incident, the subsequent programs undertaken by the industry under the leadership of INPO and NUMARC, the NRC's actions, and the overall evolution of the industry's vision in relation to this problem. This paper presents the history of the design basis efforts from the first recognition of the problem by the NRC after the Davis-Besse incident, describes the actions taken by the NRC, INPO, NUMARC, the U.S. utilities and the NSSS designers, and brings the problem statement up-to-date in relation to the vision presently held by the U.S. nuclear industry. It then presents a technical discussion to develop a detailed definition of design basis information to support the problem statement. The information originally supplied by the NSSS designers during the plant design and construction is discussed as well as its relationship to the previously defined design basis information. This section of the paper concludes by defining the additional information needed by nuclear utilities to satisfy the requirements developed from the problem statement. Having developed a definition of the additional information (i.e., information not originally supplied during design and construction) required to solve the design basis problem as it is presently perceived by the U.S. nuclear industry, the paper then discusses design basis

  11. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  12. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  13. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  14. Progress in Developing a High-Availability Advanced Tokamak Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, T.; Goldston, R.; Kessel, C.; Neilson, G.; Menard, J.; Prager, S.; Scott, S.; Titus, P.; Zarnstorff, M., E-mail: tbrown@pppl.gov [Princeton University, Princeton Plasma Physics Laboratory, Princeton (United States); Costley, A. [Henley on Thames (United Kingdom); El-Guebaly, L. [University of Wisconsin, Madison (United States); Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Waganer, L. [St. Louis (United States)

    2012-09-15

    Full text: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. In this paper the AT pilot plant design will be presented. The selected maintenance approach, the device arrangement and sizing of the in-vessel components and details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will be discussed. Efforts made to enhance the interaction of in-vessel maintenance activities, the hot cell and the transfer process to develop simplifying solutions will also be addressed. (author)

  15. Sequential Design of Experiments to Maximize Learning from Carbon Capture Pilot Plant Testing

    Energy Technology Data Exchange (ETDEWEB)

    Soepyan, Frits B.; Morgan, Joshua C.; Omell, Benjamin P.; Zamarripa-Perez, Miguel A.; Matuszewski, Michael S.; Miller, David C.

    2018-02-06

    Pilot plant test campaigns can be expensive and time-consuming. Therefore, it is of interest to maximize the amount of learning and the efficiency of the test campaign given the limited number of experiments that can be conducted. This work investigates the use of sequential design of experiments (SDOE) to overcome these challenges by demonstrating its usefulness for a recent solvent-based CO2 capture plant test campaign. Unlike traditional design of experiments methods, SDOE regularly uses information from ongoing experiments to determine the optimum locations in the design space for subsequent runs within the same experiment. However, there are challenges that need to be addressed, including reducing the high computational burden to efficiently update the model, and the need to incorporate the methodology into a computational tool. We address these challenges by applying SDOE in combination with a software tool, the Framework for Optimization, Quantification of Uncertainty and Surrogates (FOQUS) (Miller et al., 2014a, 2016, 2017). The results of applying SDOE on a pilot plant test campaign for CO2 capture suggests that relative to traditional design of experiments methods, SDOE can more effectively reduce the uncertainty of the model, thus decreasing technical risk. Future work includes integrating SDOE into FOQUS and using SDOE to support additional large-scale pilot plant test campaigns.

  16. Irradiation pilot plants and experimental facilities available for food preservation

    International Nuclear Information System (INIS)

    1975-01-01

    With the ever-increasing world food crisis mankind has to face today, the prevention of spoilage of perishable food is gaining in momentum. The World Food Conference (Rome, November 1974) of the United Nations clearly recognized the importance of food preservation and urged action in this field. Irradiation is one of the recently discovered methods to preserve food. Its practical introduction largely depends on three main factors: (a) proof of the safety for human consumption of the irradiated product, (b) technological feasibility and (c) economic competitiveness of the process. As data on safety for consumption ('wholesomeness') continue to become available, the number of countries authorizing the irradiation of certain food items is growing (present total: 17 countries), and the same is true for the number of licensed irradiated commodities (total: 23). Under these conditions, testing of the technological and economic feasibility of food irradiation is a matter of increasing importance. Economic feasibility of any industrial operation can only be studied in larger-scale experiments. Thus, they can only be performed with radiation sources larger than those found in laboratories, i.e. in pilot irradiators, capable of handling from a few hundred to a few thousand kilograms of material within a short period of time. The Food Preservation Section of the Joint FAO/IAEA Division of Atomic Energy in Food and Agriculture has attempted to collect data on the availability, for food preservation, of suitable irradiators in Member States

  17. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  18. Biofuels 2.0 move to pilot plant

    International Nuclear Information System (INIS)

    Dupin, L.

    2010-01-01

    The second generation of biofuels, which use the non-energy parts of plants, do not compete with the food industry. These biofuels have been tried and tested at the laboratory but challenges are occurring with the transition to industrial plants. Demonstrators and prototypes are developing in Germany, Japan, USA and France and bet on two different processes, the biochemical way (enzymatic reaction and fermentation) and the thermochemical way (gasification and Fischer-Tropsch synthesis). Research is in progress on a possible third generation of biofuels which will use micro-algae. The interest of this third way is triple: no competition with the food industry, no land use (production in bioreactors), and enhanced CO 2 capture. (J.S.)

  19. Socioeconomic study for the proposed waste isolation pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    This document presents the historical and existing socioeconomic conditions in the vicinity of the proposed plant, projected changes in those conditions with and without the plant, and an outline of the various techniques used to make these projections. The analysis predicts impacts on the general economy in the area near the plant and on employment, personal income, population, social structure, the private economic sector, housing, land use, community services and facilities, and local government finances. Among the most important results are the following predictions: The economy of the area will derive $165 million directly and indirectly during the first 7.5 years of the project. After that, it will derive about $21 million directly and indirectly during each year of full operation. About 2100 jobs will be created directly and indirectly at the peak of the construction and about 950 jobs during the full operation. A net in-migration will occur: about 2250 people at the peak of the construction and about 1000 people during operation. A housing shortage may begin in Carlsbad in 1981 or 1982 and last for about 2 years.

  20. Socioeconomic study for the proposed waste isolation pilot plant

    International Nuclear Information System (INIS)

    1980-10-01

    This document presents the historical and existing socioeconomic conditions in the vicinity of the proposed plant, projected changes in those conditions with and without the plant, and an outline of the various techniques used to make these projections. The analysis predicts impacts on the general economy in the area near the plant and on employment, personal income, population, social structure, the private economic sector, housing, land use, community services and facilities, and local government finances. Among the most important results are the following predictions: The economy of the area will derive $165 million directly and indirectly during the first 7.5 years of the project. After that, it will derive about $21 million directly and indirectly during each year of full operation. About 2100 jobs will be created directly and indirectly at the peak of the construction and about 950 jobs during the full operation. A net in-migration will occur: about 2250 people at the peak of the construction and about 1000 people during operation. A housing shortage may begin in Carlsbad in 1981 or 1982 and last for about 2 years

  1. Results of the DIOS pilot plant test and summary of the joint research; DIOS pilot plant no shiken sogyo kekka to kenkyu seika no matome

    Energy Technology Data Exchange (ETDEWEB)

    Sugiyama, T [Center for Coal Utilization, Japan, Tokyo (Japan); Kawaoka, K [The Japan Iron and Steel Federation, Tokyo (Japan)

    1996-09-01

    A joint research had been carried out with a subsidy from the Agency of Natural Resources and Energy since fiscal 1988 to fiscal 1995 on the direct iron ore smelting reduction process (DIOS process). The process utilizes coal directly as a process to use the strong points and supplement the weak points of the blast furnace process. During the period, a pilot plant had been operated since 1993. Upon having completed the feasibility study, this paper reports the result thereof. The main facilities consist of a smelting and reducing furnace of iron bath type, a spare reducing furnace of fluidized bed type, and a preheating furnace. The former two furnaces constitute a unit structure with the two furnaces connected vertically. The pilot plant achieved a three-day continuous operation producing 500 tons of iron every day. The production rate reached 21 tons an hour at an upward oxygen blowing velocity of about 13,000 Nm {sup 3} per hour. The coal unit requirement showed a result of <1000 kg/t for high VM coal and <900 kg/t for low VM coal. These results verified a possibility that this process can supplement or replace the blast furnace process even for a production scale of 9000 tons a day. 7 refs., 15 figs., 3 tabs.

  2. Development of pilot model of virtual nuclear power plant and its application to radiation management

    International Nuclear Information System (INIS)

    Kang, K. D.; Sin, S. W.

    2002-01-01

    Using Virtual Reality (VR) technique, a real model for radiation controlled area in nuclear power plant was developed and a feasibility study to develop a computational program to estimate radiation dose was performed. For this purpose a pilot model with an dynamic function and bi-directional communication was developed. This model was enhanced from the existing 3-D single-directional communication. In this pilot model, a plant visitor needs a series of security checking process initially. If he(she) enters the controlled area and approaches radiation hazard area, the alarms with warning lamp will be initiated automatically. Throughout the test to connect this model from both domestic and international sites in various time zones it has proven that it showed a sufficient performance. Therefore this model can be applied to broad fields as radiation protection procedures photographic data, on-line dose program

  3. Waste Isolation Pilot Plant Site Environmental Report for calendar year 1989

    International Nuclear Information System (INIS)

    1989-01-01

    This is the 1989 Site Environmental Report (SER) for the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP is a government owned and contractor-operated facility. The WIPP project is operated by Westinghouse Electric Corporation for the US Department of Energy (DOE). The mission of the WIPP is to provide a research and development facility to demonstrate the safe disposal of transuranic (TRU) waste generated by the defense activities of the US Government. This report provides a comprehensive description of environmental activities at the WIPP during calendar year 1989. The WIPP facility will not receive waste until all concerns affecting opening the WIPP are addressed to the satisfaction of the Secretary of Energy. Therefore, this report describes the status of the preoperational activities of the Radiological Environmental Surveillance (RES) program, which are outlined in the Radiological Baseline Program for the Waste Isolation Pilot Plant (WTSD-TME-057). 72 refs., 13 figs., 20 tabs

  4. Separation of packaging plastics by froth flotation in a continuous pilot plant

    International Nuclear Information System (INIS)

    Carvalho, Teresa; Durao, Fernando; Ferreira, Celia

    2010-01-01

    The objective of the research was to apply froth flotation to separate post-consumer PET (Polyethylene Terephthalate) from other packaging plastics with similar density, in a continuously operated pilot plant. A representative sample composed of 85% PET, 2.5% PVC (Polyvinyl Chloride) and 11.9% PS (Polystyrene) was subjected to a combination of alkaline treatment and surfactant adsorption followed by froth flotation. A mineral processing pilot plant, owned by a Portuguese mining company, was adapted for this purpose. The experimentation showed that it is possible to produce an almost pure concentrate of PET, containing 83% of the PET in feed, in a single bank of mechanical flotation cells. The concentrate grade attained was 97.2% PET, 1.1% PVC and 1.1% PS. By simulation it was shown that the Portuguese recycling industry specifications can be attained if one cleaning and one scavenger stages are added to the circuit.

  5. BWR alloy 182 stress Corrosion Cracking Experience

    International Nuclear Information System (INIS)

    Horn, R.M.; Hickling, J.

    2002-01-01

    Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

  6. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  7. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  8. Fiscal 1997 report of the development of high efficiency waste power generation technology. No.2 volume. Pilot plant verification test; Kokoritsu haikibutsu hatsuden gijutsu kaihatsu (pilot plant jissho shiken). 1997 nendo hokokusho (daini bunsatsu)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    As to a high efficiency waste power generation system using general waste as fuel, the details of the following were described: design/construction management and operational study of pilot plant, design/manufacture/construction of pilot plant, and study of an optimal total system. Concerning the construction management and operational study, the paper described the application for governmental/official inspection procedures and taking inspection, process management of pilot plant, site patrol, safety management, management of trial run of pilot plant, drawing-up of a verification test plan and test run, etc. Relating to the design/manufacture/construction of pilot plant, an outline of the pilot plant was described. The paper also stated points to be considered in design of furnace structure and boiler structure, points to be considered of the verification test, etc. As to the study of an optimal total system, the following were described: survey of waste gasification/slagging power generation technology, basic study on RDF production process, survey of trends of waste power generation technology in the U.S., etc. 52 refs., 149 figs., 121 tabs.

  9. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  10. Preoperational checkout of the remote-handled transuranic waste handling at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1987-09-01

    This plan describes the preoperational checkout for handling Remote-Handled Transuranic (RH-TRU) Wastes from their receipt at the Waste Isolation Pilot Plant (WIPP) to their emplacement underground. This plan identifies the handling operations to be performed, personnel groups responsible for executing these operations, and required equipment items. In addition, this plan describes the quality assurance that will be exercised throughout the checkout, and finally, it establishes criteria by which to measure the success of the checkout. 7 refs., 5 figs

  11. The Planning of a Small Pilot Plant for Development Work on Aqueous Reprocessing of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeborg, T U; Haeffner, E; Hultgren, Aa

    1963-10-15

    A shielded volume (42 m{sup 3}) in the hot laboratory at Kjeller, Norway, has been used for the installation of a small pilot plant intended for studies on nuclear fuel reprocessing. During the first period of operation (1963) a plutonium separation method (the Silex process) developed at AB Atomenergi will be studied. This document is a description of the project during the stage of technical planning and chemical process development.

  12. Resource Conservation and Recovery Act Part B permit application [for the Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    1993-01-01

    This volume contains Appendix D2, engineering design basis reports. Contents include: Design considerations for the waste hoist of the Waste Isolation Pilot Plant (WIPP); A site-specific study of wind and tornado probabilities at the WIPP Site in southeast New Mexico; Seismic evaluation report of underground facilities; and calculations for analysis of wind loads and tornado loads for WHB, seismic calculations, calculations for VOC-10 monitoring system, and for shaft at station A

  13. Conceptual design of SO3 decomposer for thermo-chemical iodine-sulfur process pilot plant

    International Nuclear Information System (INIS)

    Akihiro Kanagawa; Seiji Kasahara; Atsuhiko Terada; Shinji Kubo; Ryutaro Hino; Yoshiyuki Kawahara; Masaharu Watabe; Hiroshi Fukui; Kazuo Ishino; Toshio Takahashi

    2005-01-01

    Thermo-chemical water-splitting cycle is a method to make an effective use of the high temperature nuclear heat for hydrogen production. Japan Atomic Energy Research Institute (JAERI) has been conducting R and D on HTGR and also on thermo-chemical hydrogen production by using a thermo-chemical iodine-sulfur cycle (IS process). Based on the test results and know-how obtained through a bench-scale tests of hydrogen production of about 30 NL/hr, JAERI has a plan to construct a pilot test plant heated by high temperature helium gas, which has a hydrogen production performance of 30 Nm 3 /hr and will be operated under the high pressure up to 2 MPa. One of the key components of the pilot test plant is a SO 3 decomposer under high temperature conditions up to 850 degree C and high pressure up to 2 MPa. In this paper, a concept of the SO 3 decomposer for the pilot test plant fabricated with SiC ceramics, a corrosion-resistant material is investigated. Preliminary analyses on temperature and flow-rate distributions in the SO 3 decomposer and on thermal stress were carried out. A SO 3 decomposer model was experimentally manufactured. (authors)

  14. 1974 conceptual design description of a bedded salt pilot plant in southeast New Mexico

    International Nuclear Information System (INIS)

    1977-06-01

    The policy of the United States Atomic Energy Commission is to take custody of all commercial high-level radioactive wastes and maintain control of them in perpetuity. This policy (Title 10, Code of Federal Regulations, Part 50, Appendix F) requires that the high-level wastes from nuclear fuels reprocessing plants be solidified within five years after reprocessing and then shipped to a federal repository within ten years after reprocessing. Ultimate disposal sites and/or methods have not yet been selected and are not expected to be ready when waste deliveries begin about 1983. Therefore, the AEC plans to build an interim storage facility, called Retrievable Surface Storage Facility (RSSF), to store and isolate the waste from man and his environment until the suitability of the permanent repository is demonstrated and public acceptance has been established. Meantime, the AEC is proceeding with the study and development of an ultimate disposal method. Bedded salt is being considered for ultimate waste disposal, and work is in progress to develop a Bedded Salt Pilot Plant to demonstrate its acceptability. The pilot plant will permit in situ verification of laboratory work on the interaction of heat and radioactivity of the waste with the salt and surroundings. One concept of such a pilot facility is described

  15. [Yield of starch extraction from plantain (Musa paradisiaca). Pilot plant study].

    Science.gov (United States)

    Flores-Gorosquera, Emigdia; García-Suárez, Francisco J; Flores-Huicochea, Emmanuel; Núñez-Santiago, María C; González-Soto, Rosalia A; Bello-Pérez, Luis A

    2004-01-01

    In México, the banana (Musa paradisiaca) is cooked (boiling or deep frying) before being eaten, but the consumption is not very popular and a big quantity of the product is lost after harvesting. The unripe plantain has a high level of starch and due to this the use of banana can be diversified as raw material for starch isolation. The objective of this work was to study the starch yield at pilot plant scale. Experiments at laboratory scale were carried out using the pulp with citric acid to 0,3 % (antioxidant), in order to evaluate the different unitary operations of the process. The starch yield, based on starch presence in the pulp that can be isolated, were between 76 and 86 %, and the values at pilot plant scale were between 63 and 71 %, in different lots of banana fruit. Starch yield values were similar among the diverse lots, showing that the process is reproducible. The lower values of starch recovery at pilot plant scale are due to the loss during sieving operations; however, the amount of starch recovery is good.

  16. Future-oriented computerized information system for power plant process control in a pilot project at Philippsburg nuclear power plant

    International Nuclear Information System (INIS)

    Woehrle, G.; Kraft, M.

    1988-01-01

    The motivation for the pilot project at Philippsburg nuclear power plant resulted from the Three Mile Island accident in 1979. The primary task embraces an efficient computer-aided reduction of information when a fault occurs based on a process engineering analysis of the information accrued. Accompanying this are a consolidation and evaluation of the information available in the control room. In this pilot project the new tasks of status monitoring, information reduction and operationalcontrol have been realized for the first time using a computer-aided process information system. In addition to the existing control computer, an information computer with approximately 1200 analogue and about 10000 binary signals has been installed. The installation of the system was completed in 1984 and in the meantime initial operational experience has become available. (orig.) [de

  17. BWR reactor management system

    International Nuclear Information System (INIS)

    Makino, Kakuji; Kawamura, Atsuo; Yoshioka, Ritsuo; Neda, Toshikatsu.

    1979-01-01

    It is necessary to grasp the delicate state of operation in reactor cores in view of the control of burn-up and power output at the time of the operation management of BWRs. Enormous labor has been required for the collection, processing and evaluation of the data. It is desirable to obtain the safer, more efficient and faster method of operation control by predicting the states in cores including the change of xenon and reflecting them to operation plans as well as by tracing with high accuracy the past burn-up history for a long period. At present, the on-line evaluation of the states in cores is carried out with the process computers attached to respective units, but the amount of data required for core operation management of high degree far exceeds their capacity. From such viewpoints, the research and development on the reactor management system were carried out. The data processing concerning core operation management is performed with newly installed computers utilizing the data from existing process computers, and the operation of reactor cores, the qualitative improvement of management works, labor saving, and fast, efficient operation control are feasible with it. This system was installed in an actual plant in October, 1977. The composition of the system, the prediction of the change in local output distribution accompanying control rod operation, the prediction of the change in the states in cores due to the flow rate of coolant, and the function of collecting plant data are explained. (Kako, I.)

  18. BWR chromium chemistry

    International Nuclear Information System (INIS)

    Baston, V.F.; Indig, M.E.; Skarpelos, J.M.

    1992-10-01

    This report addresses the concern about higher total specific conductivity in the reactor recirculation loop water due to the chromate ion. This concern is particularly high at plants where all other ionic species have been reduced through careful attention to makeup and condensate polisher operations. An EPRI Chromate Workshop was held in November 1990 to consider the issues raised by observed levels of chromate ion (generally 5 to 50 ppB). While BWRs on normal water chemistry were the only ones observing chromate, even plants on hydrogen water chemistry (HWC) observe sharp spikes of conductivity due to chromate whenever the hydrogen supply was interrupted after a reasonably long HWC operational period. The consensus of the workshop attendees was that chromate was not a concern as an agent causing pipe cracking compared to the more common species such as chloride and sulfate. However, the data are somewhat ambiguous for levels of chromate above 50 ppB. Adjustments to the weighing factors for the various ionic species in the industry chemistry performance index are suggested to allow for the known relative higher aggressiveness of other species relative to that of chromate

  19. LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    Zidan, W.I.

    2000-01-01

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  20. Textile wastewater treatment and reuse by solar catalysis: results from a pilot plant in Tunisia.

    Science.gov (United States)

    Bousselmi, L; Geissen, S U; Schroeder, H

    2004-01-01

    Based on results from bench-scale flow-film-reactors (FFR) and aerated cascade photoreactors, a solar catalytic pilot plant has been built at the site of a textile factory. This plant has an illuminated surface area of 50 m2 and is designed for the treatment of 1 m3 h(-1) of wastewater. The preliminary results are presented and compared with a bench-scale FFR using textile wastewater and dichloroacetic acid. Equivalent degradation kinetics were obtained and it was demonstrated that the solar catalytic technology is able to remove recalcitrant compounds and color. However, on-site optimization is still necessary for wastewater reuse and for an economic application.

  1. Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

    Energy Technology Data Exchange (ETDEWEB)

    Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

    1972-11-01

    In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

  2. Radiation treatment of sewage sludge - experience with an operating pilot plant

    International Nuclear Information System (INIS)

    Suess, A.; Lessel, T.

    1977-01-01

    After an operation time of a pilot plant for the γ-irradiation of sewage sludge after 3 years promising results could be obtained for economic considerations, killing rate of pathogenes and radiation induced changes in sedimentation properties. Irradiated sewage sludge indicated nearly the same effect on soil and plant as untreated. No special trained personnel are necessary for maintenance because of the simple design. Successful experience during 18 months resulted in an increase of the daily capacity up to 120 m 3 from December 1975. (author)

  3. Some economic considerations for a pilot plant based on the adsorber loop concept

    International Nuclear Information System (INIS)

    Koske, P.H.; Ohlrogge, K.; Jager, W.

    1984-01-01

    Some first order estimates about the production costs of seawater uranium are presented on the basis of a pilot plant with a capacity of about 100 t uranium per year. The plant is assumed to be operating at high seas using the ''adsorber loop concept'' for the contact between seawater and adsorber granulate. The effects of some process relevant components and parameters are discussed with regard to their contribution to the production costs and in order to analyse their cost-cutting potential. (orig.) [de

  4. Experimental Investigation and Modelling of a Wet Flue Gas Desulphurisation Pilot Plant

    DEFF Research Database (Denmark)

    Kiil, Søren; Michelsen, Michael Locht; Dam-Johansen, Kim

    1998-01-01

    A detailed model for a wet flue gas desulphurisation (FGD) pilot plant, based on the packed tower concept, has been developed. All important rate determining steps, absorption of SO2, oxidation of HSO3-, dissolution of limestone, and crystallisation of gypsum were included. Population balance...... equations, governing the description of particle size distributions of limestone in the plant, were derived. Model predictions were compared to experimental data such as gas phase concentration profiles of SO2, slurry pH-profiles, solids content of the slurry, liquid phase concentrations, and residual...

  5. Pilot incineration plant for solid, combustible, and low-level wastes

    International Nuclear Information System (INIS)

    Francioni, W.M.

    Radioactively contaminated wastes are formed in the handling of radioactive materials at the Federal Institute for Reactor Research (FIRR) and in other facilities, hospitals, sanitoria, industry, and nuclear power plants. A large part of the wastes are combustible and only very slightly radioactive. Incineration of these wastes is obvious. A pilot incineration plant, henceforth called the PIP, for radioactive combustible wastes of the FIRR is surveyed. The plant and its individual components are described. The production costs of the plant and experience gained in operation available at present are reviewed. Solid combustible radioactive waste can be incinerated in the PIP. The maximum possible reduction in volume of these wastes is achieved by incineration. Subsequently the chemically sterile ashes can be consolidated in a stable block suitable for long-term storage mixing with cement

  6. Optimisation of a wet FGD pilot plant using fine limestone and organic acids

    DEFF Research Database (Denmark)

    Frandsen, Jan; Kiil, Søren; Johnsson, Jan Erik

    2001-01-01

    , but the residual limestone content in the gypsum increased to somewhere between 19 and 30 wt%, making this pH range unsuitable for use in a full-scale plant. The investigations have shown that both the addition of organic acids and the use of a limestone with a fine PSD can be used to optimise wet FGD plants. (C......The effects of adding an organic acid or using a limestone with a fine particle size distribution (PSD) have been examined in a wet flue gas desulphurisation (FGD) pilot plant. Optimisation of the plant with respect to the degree of desulphurisation and the residual limestone content of the gypsum...... has been the aim of the work. In contrast to earlier investigations with organic acids, all essential process parameters (i.e. gas phase concentration profiles of SO(2), slurry pH profiles. and residual limestone in the gypsum) were considered. Slurry concentrations of adipic acid in the range of 0...

  7. Pilot Plant for treating of spent exchange resins

    International Nuclear Information System (INIS)

    Iglesias, Alberto M.; Raffo Calderon, Maria del C.; Varani, Jose L.

    2004-01-01

    Spent exchange resins that have been accumulating during the last operational 30 years in Atucha I nuclear power plant (NPP) are a 'problematic' waste. These spent resins conform an intermediate level waste due to the total content of alpha, beta and gamma emitters (some samples of spent resins were analyzed in 2003). For this reason its treatment is more expensive since it is necessary to add more safety barriers for its final disposition and also for the radioprotection actions that are involved. Using sulfuric acid solutions it is possible to elute from the spent resins the ions that are retained. In the same operation are eluted Cobalt, Cesium and alpha emitters since that all these elements react as cations in aqueous solution. Decontamination by electrochemical methods was analyzed as an interesting method to apply after elution operation to these spent resins since that with the decontamination process it is possible to obtain a solid without activity and concentrate the activity in cells that are small in volume and its manipulation doesn't present any extra complication. Experiments made with active samples taken from the deposit were successful. Because of these results it was built a small plant to treat a batch of 100 dm 3 of wet spent exchange resins. Some problems with the material that was in the deposit together with spent resins caused that we had to plan a more complex strategy to obtain a complete decontamination of the spent resins (in this stage we used the cobalt retention cell that was described in other paper to retain Cobalt and alpha emitters and a sample of zeolites from Argentina ores to retain Cesium). Due to alpha emitters act electrochemically like cations it was possible to retain altogether with ionic Cobalt on the copper amalgam electrode. Working in the non-active lab with alcoholic solutions it was possible to retain ionic Cesium on a copper electrode (copper is covered by mercury fine film which forms a solid amalgam) with a

  8. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices III and IV. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    The items listed below summarize the detail sections which follow: a listing of definitions and a discussion of the general treatment of data within the random variable approach as utilized by the study; a tabulation of the assessed data base containing failure classifications, final assessed ranges utilized in quantification and reference source values considered in determining the ranges; a discussion of nuclear power plant experience that was used to validate the data assessment by testing its applicability as well as to check on the adequacy of the model to incorporate typical real incidents; an expanded presentation of the data assessment giving information on applicability considerations; a discussion of test and maintenance data including comparisons of models with experience data; and special topics, including assessments required for the initiating event probabilities and human error data and modeling.

  9. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  10. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  11. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  12. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

  13. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  14. BWR Servicing and Refueling Improvement Program: Phase I summary report

    International Nuclear Information System (INIS)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

  15. FY 1991 report on the results of the development of an entrained bed coal gasification power plant. Part 4. Operation of pilot plant; 1991 nendo seika hokokusho. Funryusho sekitan gaska hatsuden plant kaihatsu - Sono 4. Pilot plant unten sosa hen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-01-01

    A record was summarized of the operation of the 200 t/d entrained bed coal gasification pilot plant that was constructed with the aim of establishing technology of the integrated coal gasification combined cycle power generation. As to the actual results of operation hours, the paper summarized the records of gasifier facilities, gas refining facilities, gas turbine facilities and safety environment facilities which were collected from April 1991 to January 1993. Relating to the actual results of start-up/stop, the paper summarized the records of gasifier facilities, gas refining facilities (desulfurization), gas refining facilities (dedusting), gas turbine facilities and safety environment facilities. Further, operation manuals were made for the schedule of plant start-up/stop, generalization, gasifier facilities, gas refining facilities (desulfurization), gas refining facilities (dedusting), gas turbine facilities, actual pressure/actual size combustor testing facilities and safety environment facilities. (NEDO)

  16. PILOT PLANT STUDY ON NATURAL WATER COAGULANTS AS COAGULAN AIDS FOR WATER SUPPLY

    Directory of Open Access Journals (Sweden)

    B BINA

    2001-06-01

    Full Text Available Introduction: Natural plant coagulants have an important role to play in provision of portable water to rural communities in the developing world. The plant material that their coagulation properties have been confirmed in previous lab scale studies and can be found widely in Iran was selected as coagulant aids. Pilot plant study was done to evaluate the efficiency of natural material such as Starch/Gum Tragacanth, Fenugreek and Yeast as coagulant aids in conjunction with comercial alum. Methods: The pilot was placed in Isfahan Water Treatment Plant (IWTP and efficiency of these materials in removal of turbidity from raw water enters the IWTP was evaluated. The results indicated while these materials were used as coagulant aids in concentration of 1-5 mg/l conjunction with alum are able to reduced the turbidity and final residuals turbidity meets the standards limits. Results: The coagulation efficiency of these material were found to be effected by certain physico-chemical factors, namely, concentration of suspended solids, divalent cation metal and time of agitation. The relative importance of these variable was evaluated. The results of COD test proved that the natural coagulant aids in the optimum doses produce no any significant organic residual. Discussion: Economical considerations showed that using of these material as coagulant aids can cause reduction in alum consumption and in some cases are more econmical than synthetic polyelectrolyte.

  17. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  18. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  19. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  20. Reducing BWR O and M costs through on-line performance monitoring

    International Nuclear Information System (INIS)

    Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

    1995-01-01

    Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants