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Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant

2010;():1-6. doi:10.1115/ICONE18-29021.

The solution to a Markov chain modeling electric power supply to critical equipment in a typical 4-loop pressurized water reactor following a Loss of offsite power event is compared with a convolution method. The standard “convolution integral” approach is described, and an alternative methodology based on a Markov model is illustrated.

Topics: Modeling
Commentary by Dr. Valentin Fuster
2010;():7-16. doi:10.1115/ICONE18-29036.

The current fleet of 104 nuclear power plants in the U.S. began their operation with 40 years operating licenses. About half of these plants have their licenses renewed to 60 years and most of the remaining plants are anticipated to pursue license extension to 60 years. With the superior performance of the current fleet and formidable costs of building new nuclear power plants, there has been significant interest to extend the lifetime of the current fleet even further from 60 years to 80 years. This paper addresses some of the key long term technical challenges and identifies R&D needs related to the long term safe and economic operation of the current fleet.

Commentary by Dr. Valentin Fuster
2010;():17-19. doi:10.1115/ICONE18-29041.

The power generation capability of Unit 2 of Ling Ao Nuclear Power Plant has been decreased about 6MW after its fifth refueling outage in 2008. This paper investigates the activities which may influence the unit capability in the outage, compares the related parameters before and after the outage, and then analyzes its root causes. From the root cause analysis, it can be found that the replacement of orifices which measure the feedwater flow was the main reason for the capability decrease. Finally, some modification measurements have been provided to recover the unit capability.

Commentary by Dr. Valentin Fuster
2010;():21-26. doi:10.1115/ICONE18-29045.

Dissimilar Metal Welds have been widely used at safe-related pressure vessels and piping in Nuclear Power Plants. Some industry codes have been developed for nuclear power plants, such as ASME BPVC volume III & XI, Germany KTA 3201.3 and 3201.4 code and Russia code of light water nuclear power plants. The difference of those codes and some industry feed backs and some experiment results have been briefly introduced and discussed. Furthermore, the inspection qualification or performance demonstration, one of new requirements of some codes, especially for pre-service and inservice inspection of nuclear power plants, really promote continual improvements of DMWs UT both on detection rate and sizing accuracy. Additional, these works really benefit the revision of EJ/T 1039-1996, Non-destructive testing for mechanical components in nuclear island of nuclear power plants, as china building more and more nuclear power plants.

Commentary by Dr. Valentin Fuster
2010;():27-32. doi:10.1115/ICONE18-29051.

The capacity of the electrical power system in Egypt will increase rapidly in the coming twenty years. In year 2018, nuclear power generation will be connecting to the Egyptian electrical grid. Consequently, the interaction of nuclear power plants and other systems becomes a very important issue, and a detailed nuclear power model for the medium-term and long-term power system stability should be developed. However, there is no nuclear unit model that can describe the detailed characteristics of the nuclear unit in the available commercial power system simulation software. In this paper, a detailed pressurized water reactor (PWR) nuclear unit model for medium-term and long-term power system transient stability is proposed. The model is implemented by a user defined program in PSS/E through PSS/E Matlab Simulink Interface. This model can be used to analyze the interaction of nuclear power plants and other power systems. The simulation results show that the proposed model is valid.

Commentary by Dr. Valentin Fuster
2010;():33-39. doi:10.1115/ICONE18-29095.

10 MW High Temperature Gas-cooled Reactor (HTR-10) is a pebble bed reactor, helium serves as the coolant, the fuel element is spherical, and its diameter is 60 mm. In power operation, the reactivity of HTR-10 is almost kept in a constant, because the fuel elements could be loaded, discharged and reloaded online, all these processes referred above are carried out automatically by the fuel handling system. In this paper, the operation characteristics of the fuel handling system, including the features of some special equipments, three automatic processes of fuel elements handling, the configuration and operation features of the control system and the operation conclusion in these years since 2000 are introduced.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2010;():41-45. doi:10.1115/ICONE18-29096.

There are three research reactors of different type in Institute of Nuclear and New Energy Technology of Tsinghua University (INET), they are Swimming Pool Shielding Reactor (SPSR), 5MW Nuclear Heating Reactor (NHR-5) and 10MW High Temperature Gas Cooled Reactor (HTR-10). The interim storage facilities for Spent Nuclear Fuel (SNF) are applied in these reactors except NHR-5, the SPSR adopts the wet storage for its SNFs, HTR-10 accepts the dry storage due to the ceramic structure of its SNFs. The practical storage conditions including SNFs features, space distribution, SNFs transportation, shielding measures and the safety analysis results including radioactive activity, shielding effect, the total amount of radioactive product leakage to environment are demonstrated in this paper.

Topics: Storage
Commentary by Dr. Valentin Fuster
2010;():47-52. doi:10.1115/ICONE18-29133.

The displacements of nuclear island buildings occur because of earthquake, thermal expansion, creep shrinkage etc. Because there are penetrations and some supports that connect to the wall of reactor building, the displacement loads must be considered when the piping and the supports are verified, the displacement loads have great effect on the piping and supports stress analysis. Studying the effect and looking for that how to reduce this effect are helpful to find a way to reduce the piping stress and the loads on the supports, and it is also helpful to make the plant safe. In this paper some piping calculation units of EPR project are showed as examples, the calculation units are done with a new software ANASYS PIPE, and finite element method is used. there are penetrations and supports that connects to the wall of reactor building in the calculation units, the building displacements due to earthquake, thermal expansion and creep shrinkage are considered, and the effect of displacement loads and the combinations of all kinds of loads under different category conditions on the piping and supports are studied, the result shows that the stress of piping and the loads on the supports caused by building displacement loads are prominent, but the piping and the supports can meet the requirement by optimizing the support concept.

Commentary by Dr. Valentin Fuster
2010;():53-58. doi:10.1115/ICONE18-29136.

This paper describes the general practice and lessons learned from Commercial-Grade Item dedication for nuclear safety-related system applications in Taiwan. The dedication process qualified the commercial off-the-shelf components to be applied as basic components. In past fifteen years, Institute of Nuclear Energy Research (INER) has actively performed the dedication service to help local nuclear power plants solve their procurement problems of nuclear grade items, due to reduced availability of qualified suppliers and/or obsolete issues of qualified components. The Scope of dedication includes material, electrical and mechanical components located in mild and harsh environment. Thousands of components such as piping, fitting, breaker, relay, motor, and control device etc., have already been dedicated to and successfully used in local nuclear power plants. The Commercial-Grade Item dedication process is based on EPRI documentations. Besides, the technical evaluation and equipment qualification are included during the dedication process. The requirements for equipment qualification are described in IEEE standards. Although the codes and standards for dedication in Taiwan refer to those in USA, the challenges may happen due to different regulators, utility, manufacture’s quality culture, and personal responsibility. The key to successful dedication will be dependent on the well-defined component requirements and good project planning. This paper introduces the self-reliant experiences in dedication and economic benefit to local nuclear power plants.

Topics: Safety
Commentary by Dr. Valentin Fuster
2010;():59-65. doi:10.1115/ICONE18-29139.

The 10 MW high temperature gas-cooled reactor–test module (HTR-10) reached the first critical at the end of year 2000, and has been running for over 9 years safely and stably till now. Comparing with the Pressure Water Reactor (PWR), HTR-10 has many different characteristics, such as core construction, special fuel elements, helium coolant and so on. Thus the thermal hydraulic parameter measurement has special requirement and it is indispensable to select or develop some new class 1E instrumentation and devices. This paper describes measurement requirements, measurement method and measurement instrumentations for measuring coolant temperature, primary loop pressure, primary loop mass flow rate, primary loop humidity, main steam pressure, feedwater mass flow rate, in-core components temperature, pressure vessel surface temperature. The class 1E sheathed thermocouples and thermocouple penetration assembly, the class 1E orifice plate throttle device, and the data acquisition and supervision system that were developed by the Institute of Nuclear and New Energy Technology (INET) according to the criteria IEEE 323 and IEEE 344 are introduced in detail. HTR-10 has been operated successfully for over 9 years up to the present. The operation and maintenance experience of above-mentioned instrumentations shows they are safe and reliable at normal and abnormal conditions. The experience described in this paper is valuable for the latter 2 × 250 MW modular high temperature gas-cooled reactor.

Commentary by Dr. Valentin Fuster
2010;():67-72. doi:10.1115/ICONE18-29147.

This paper provides summary of domestic in service nuclear power plant condensate polishing system (CPS) configuration, CPS is supplied for all domestic in service nuclear power plant and almost all above sub–critical fossil power plant, so condensate polishing system is necessary for PWR nuclear reactor. Based on USA industrial practice, the raw condensate for startup and normal operation is presented in this paper. Base on EPRI PWR nuclear reactor secondary water chemistry guideline for steam generator blowdown water quality, ion and water balance in secondary system, the effluent water quality for CPS is calculated. The condensate flow is proportional with electric power, during the plant startup operation, the electric power level from 0 to 30% to 50% to 100%, is longer time, i. e. condensate flow from 0 to 30% to 50% to 100%, is longer time. Before the plant electric power level 50% (or 30%), the impurity in condensate is purified, the makeup water for secondary system is demineralied water, the impurity level is lower than requirement, so 50% (or 30%) flow condensate polishing system is enough for startup operation. During the plant normal operation, condensate polishing system is in hot standby, if the condensate tube leak, condensate is deteriorated, the condensate polishing system need to be operated immediately, but CPS put in service from hot standby at least 5 minute. During this period, the units which is once–though cooled by sea water will be shutdown immediately with 2.71/h continuous condensate tube leakage, so full flow CPS is necessary for the units which is cooled by sea water, and it is better that one or two series put in service during normal operation other than all of them in hot standby. The units which is cooled by fresh water will be shutdown immediately with 2561/h continuous condensate tube leakage, for this level leakage, whichever 50%(30%) and 100% CPS, action level 2 will be preformed for the units, so 50%(30%) is enough for fresh water cooled units.

Commentary by Dr. Valentin Fuster
2010;():73-79. doi:10.1115/ICONE18-29155.

At Hamaoka Unit 4 and 5, the hydrogen concentration in the outlet of off-gas recombiner had increased, and the reactors could not continue start-up operation. Therefore, we investigated the causes of the deactivating the recombination reaction and selected appropriate countermeasures to the plants. From our investigation, two types of deactivation mechanism are found. One of the causes was decreasing the active surface area of alumina as support material by the dehydrative condensation. The other cause was poisoning of the catalyst by organic silicon compound. The organic silicon was introduced from organosilicon sealant used at the junctions of the low-pressure turbine. We also found that the boehmite rich catalyst was deactivated more easily by the organic silicon than gamma alumina because boehmite had a lot of hydroxyl groups. Finally, we estimated that the deactivation of the hydrogen recombination catalysts was caused by combined two factors, which are characteristics of boehmite catalyst support and the poisoning by the organic silicon on the catalyst surface. As the countermeasures, the boehmite was changed into more stable gamma alumina by adding the heat treatment in hydrogen atmosphere at 500°C for 1 hour, and the source of organic silicon, organosilicon sealant, was removed. At Hamaoka Unit 4 and 5 improved catalysts were applied. Moreover, linseed oil that used to be used at the plants was applied again as sealant of the low-pressure turbine casing instead of the organosilicon sealant. As a result of application of these countermeasures, the reactors could be started without increase of the hydrogen concentration at these plants.

Topics: Catalysts , Hydrogen
Commentary by Dr. Valentin Fuster
2010;():81-90. doi:10.1115/ICONE18-29157.

The recent trend in increasing the power generation capacity of pressurized water reactors (PWRs) up to the 1700MW Class for greater economy is met with inherent challenges. Higher generation capacity necessitates turbines with higher efficiency. Operating the high pressure (HP) turbine at higher efficiency requires development of large size moisture separator reheater (MSR) to accommodate the higher specific steam volume due to the reduced HP exhaust pressure. Higher specific volume of steam also results in higher velocities and increased pressure drops. At higher steam velocity, the flow accelerated corrosion (FAC) is enhanced and the moisture separator performance will be deteriorated due to the increased mist carry over across the Chevron type vanes. The development of MSR for up to 1700MW Class PWR involved optimizing the heat balance around the MSR and selecting the optimized size of the MSR for performance and cost. This also included selection of an optimized terminal temperature difference (TTD) for the MSR. Then the newly developed MSR was verified by computational (using CFD) and experimental techniques. Using CFD, the pressure, flow, and velocity distribution, pressure drop, and velocity distribution on the shell surface were analyzed. An experimental set up using actual size sliced model of the MSR was used to predict the separator performance and drainage characteristics. The testing was carried out with two phase air-water flow at atmospheric conditions. This paper summarizes the large size MSRs for up to 1700 MW Class PWR, its development, and design verification.

Commentary by Dr. Valentin Fuster
2010;():91-96. doi:10.1115/ICONE18-29194.

Environmental protection requirement is more and more critical now, and it increases the request to prevent dangerous liquid to leak outside in nuclear power plant too. Centrifugal pumps are the most important active equipments in nuclear power plant, but there is a shaft clearance between rotor and stator of centrifugal pump. The shaft clearance can lead pumped fluid to the outside, so the environment may be polluted by the leakage. In some critical conditions such as transferring high radioactive fluid in the pump, the leakage shall be totally forbidden. So solutions have to be found to make centrifugal pumps totally leak-free for applications in nuclear power plant. Normally there are three leak-free technologies for centrifugal pumps: mechanical seal with auxiliary system, canned motor and magnetic drive. In this paper, all the three leak-free technologies and some of their applications in EPR 3rd generation PWR nuclear power plants are presented and discussed. The results show that in EPR nuclear power plant, canned motor pumps can be preferably used for strict environmental requirement of leak-free if the pump power and operating conditions are applicable. For other conditions, pumps with double mechanical seal can also be used with additional sealing water system support. For centrifugal pumps with magnetic drive are not so applicable in high pressure condition, and the safety aspect is weaker than canned motor pumps, generally they are not used in EPR nuclear power plant at present.

Commentary by Dr. Valentin Fuster
2010;():97-102. doi:10.1115/ICONE18-29277.

The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam, India. The reactor went critical in October, 1985 with a core of 23 unique high Plutonium carbide fuel subassemblies and the reactor power was rated for 10.5MWt with peak linear heat rating of fuel at 320W/cm. The extension of the target burn-up of this fuel based on Post Irradiation Examination at different stages enabled progressive expansion of the core and increase in reactor power. The reactor has been operated upto a power level of 18.6MWt/3MWe with a sodium temperature of 482°C max. The reactor has completed 24 years of operation and is currently under periodic safety review by the Atomic Energy Regulatory Board of India. As a part of the periodic safety review, equipment qualification status and ageing management studies have been presented to the regulators. Equipment qualification refers to the ability of the replaceable equipment to meet the functional requirements on demand, accomplished by periodic surveillance, maintenance and replacement. Ageing management addresses the residual life assessment of components which are passive, non-replaceable / replaceable with difficulty, taking into account their life degrading mechanisms. Over a period of time, based on the operational feedback, maintenance difficulties and obsolescence, several major components have been replaced. These include the Neutronic channels, UPS, computers of the Central Data Processing System, main boiler feed pumps, three control rod drive mechanisms, two control rods, central canal plug, deaerator lift pumps, reheaters of the steam water system, station batteries, DM plant and Nitrogen plant. The starting air system of the emergency diesel generators and isolation dampers of the reactor containment building have also been replaced. Regarding the non-replaceable components, residual life assessment has been carried out based on the operational history vis-à-vis the design limits for each component. The life limiting mechanism of heat transport systems of FBTR are creep and fatigue. Since the reactor has operated only upto a temperature of 444°C till 2007, the creep effect is insignificant. The total number of thermal cycles seen by the reactor components as of 2007 was 163, as against the design cycle of 2000 for most of the components. Hence all the heat transport system components are as good as fresh ones. However, the major life limiting factor has been found to be the Neutronic fluence on the grid plate which supports the core. The fast flux at the grid plate location was measured using Np foils and the residual life of the reactor has been assessed to be 10.5 effective full power years. This paper details the life extension exercise being carried out for FBTR.

Topics: Life extension
Commentary by Dr. Valentin Fuster
2010;():103-108. doi:10.1115/ICONE18-29316.

This paper makes a brief introduction on AP1000 operation procedure system, including procedure classification, function and composition. In addition, key points of work flow process and the advantages of AP1000 operation procedures are described, among which the application of CPS (computerized procedure system) on AP1000 operation area and human factor engineering are highlighted. CPS, as an advanced procedure system, which is relatively new to existing nuclear power plants in China, does not only have the function of electronic indication for procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. Moreover, based on current situation, this paper offers several suggestions on procedure development for Sanmen AP1000 nuclear power project, i.e. first, we can ensure the quality of operation procedures by preparing a precise writer’s guideline, a friendly-interfaced procedure template, an efficient work configuration and an appropriate schedule; then determine the way how we are going to use operation procedures in English version; finally realize CPS Chinesization and localization gradually by digesting and absorbing API 000 technology from Westinghouse Electric Company. This paper gives an intact and systematic discourse on AP1000 operation procedure system and its characteristics. Besides, the latter part of this paper focuses on development of AP1000 operation procedures for Sanmen nuclear power plant and it would be a worthwhile reference for newly-built AP1000 units in China.

Commentary by Dr. Valentin Fuster
2010;():109-111. doi:10.1115/ICONE18-29324.

A recently developed technique “Spread Spectrum Time Domain Reflectometry” (SSTDR), and supporting test devices will be adapted and tested to monitor and diagnose nuclear plant electrical systems. Current time domain reflectometry methods cannot detect or locate small faults after arc fault events, because their impedance discontinuity is too small and transient to create a measurable reflection. However, on-line, unobtrusive SSTDR can detect and locate arc and other electrical faults when the (∼msec) short circuit returns a strong reflected signal. These observations have led to development of SSTDR. If SSTDR can be successfully adapted to present and future nuclear plant electrical systems, it will be possible to monitor, on-line, the integrity of the electrical system continuously and with only minor equipment modification and no consequential safety issues. An integrated circuit (IC) is under development at the University of Utah for applications in the aircraft industry that will be adapted and used for this proposed development.

Commentary by Dr. Valentin Fuster
2010;():113-119. doi:10.1115/ICONE18-29370.

Under sodium sensor has been developed to inspect heat exchanger pipes in FBR plants. Several methods have been applied to manufacture the volumetric inspection sensor. Low temperature diffusion bonding method was selected to bond piezoelectric element and thin stainless steel plate together for the boundary between piezoelectric element and sodium. Under sodium test has been made to verify the detection ability of the newly developed volumetric inspection sensor. Under sodium tests obtained that the sensor could detect 20% depth EDM (Electro discharge method) slits in outer and inner surface of 1mm thickness pipe wall.

Commentary by Dr. Valentin Fuster
2010;():121-124. doi:10.1115/ICONE18-29378.

A set of screening and classification method is proposed in this paper to manage the Systems, Structures and Components (SSCs) of the Pressurized Water Reactor (PWR) for the aging management. The method proposed and carried out in China Nuclear Power Plants (NPP) is based on the systematic aging evaluation method and the management method for the license renewal from International Atomic Energy Agency (IAEA) and U.S. Nuclear Regulatory Commission (NRC), whose successful management methods are discussed in this paper. Considering the importance of the classification management of the SSCs, the classification method is investigated from the aspects of the safety, reliability and replacement of the SSCs with the Nuclear Power Plant operation experience feed-back. With the classification management method, the SSCs are managed economically and effectively for the Nuclear Power Plants in China. The method also makes great foundation for the aging management and residual life evaluation of the PWR in China.

Commentary by Dr. Valentin Fuster
2010;():125-130. doi:10.1115/ICONE18-29404.

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

Commentary by Dr. Valentin Fuster
2010;():131-136. doi:10.1115/ICONE18-29424.

Integrated method for constructing proactive trouble prevention knowledge base has been studied before starting the practical system development towards various application areas of knowledge base system for failure modes and effect analysis (FMEA), fault tree analysis (FTA), and so forth. In this paper, basic ideas of configuring the structured knowledge based systemization to be used for such purpose are summarized with the following order: (i) Effective re-use of various trouble information, (ii) Method of structuralizing trouble knowledge, (iii) Knowledge on trouble prediction and proactive prevention, (iv) Description of energy, mass and information flow by multilevel flow model, and (v) Method of knowledge base systemization.

Commentary by Dr. Valentin Fuster
2010;():137-144. doi:10.1115/ICONE18-29454.

Exergy analysis model of PWR nuclear power station is developed in which signal flowing graph theory is introduced to set up the relation equations between input exergy flow and output exergy flow. Then, combining with resource distribution between different components, thermo-economic analysis model is obtained by setting up unit thermo-economic cost equations of different components with productive structure graph. Taking Daya Bay as an example, exergy analysis and thermal-economic analysis are put forward with detailed distribution of exergy and investment cost. Finally, aimed at energy-saving, static diagnosis is performed in two levels: energy conservation and cost reduction, and on this basis dynamic diagnosis is developed through sensitivity analysis considering different influence factors such as main steam temperature, fuel price, construction capital investment, post treatment cost and so on. The introduction of signal flow graph theory and thermal-economic structure theory is helpful to do performance estimation with high speed and good accuracy. It provides a new way for rapid optimization and offers an effective theoretical method for energy-saving of PWR nuclear power station including advanced reactor such as AP1000.

Commentary by Dr. Valentin Fuster
2010;():145-148. doi:10.1115/ICONE18-29457.

The inspection of the steam generator divider plate becomes more and more important since indications have been detected in some steam generators in the area of the welds between tube sheet and divider plate. Remote operated manipulator and tools have been developed in order to minimize the radiation exposure of the personnel performing the inspections. Inspection techniques based on liquid penetrant, visual inspections have been developed and qualified to detect the indications. An ultrasonic inspection technique to determine the depth of the indications has been developed and qualified. The aim of this paper is to describe the inspection capabilities available to support the partition plate inspections.

Topics: Inspection , Boilers
Commentary by Dr. Valentin Fuster
2010;():149-155. doi:10.1115/ICONE18-29458.

EDF operates a fleet of 58 Pressurized Water Reactors (PWR). The “health” of the Steam Generators (SGs) is an essential element contributing to the overall thermal efficiency of a PWR, and finally to the availability of the unit. Among the health issues that may affect SGs, secondary-side corrosion products transport in PWRs may lead to many problems: various contaminants, both particulates and dissolved species, will unavoidably accumulate and concentrate in the Steam Generator. One consequence is the fouling of the heat transfer and support structure interfaces within the SG on the secondary side, especially the U-tubes (fouling deposits on the outer walls of the U-tubes), and the tube support plates (TSPs) that support the U-tubes. The accumulation of the corrosion products may lead to 3 main safety risks that must be monitored: fluid-elastic instability of tubes in flow-accelerated areas, a reduction in SG water mass inventory and an increase in the risk of water level oscillation. It has also significant performance issues because of the decision to power derate of some EDF PWRs. Thus, a global strategy to monitor the fouling and TSP blockage issues and to schedule preventive and curative actions has been designed and is under deployment by EDF nuclear operator. This dedicated periodic test relies on the recording of the following measurements in stabilized configuration: steam pressure, feedwater flowrate and temperature, primary circuit temperatures, SG blowdown flowrate and SG water level (wide and narrow range). A more precise monitoring of potential TSP blockage situations would be an interesting help to operation and maintenance strategies: deposit build-up in TSP foils could be minimized, preventive chemical cleaning operations could be scheduled and a more efficient fleet wide SG Management Program (SGMP) could be designed in accordance with secondary side deposit issues. Consequently, EDF R&D is experimenting a new method based on modeling dynamics behavior of SGs to assess a spatially distributed estimator of the TSP blockage ratio. This method, based on a 1D physical model of the SG that simulates the complex dynamics of the two-phase flow phenomena inside the SG, consists in computing the wide range water level responses according to various configurations during a particular transient which is particularly sensitive to this phenomenon. The TSP blockage ratio estimator is then obtained by comparing the computed response curves to those measured on-site. This new method has the potential advantages of being fully non-invasive, of providing a quarterly update of the TSPs blockage estimator, and of requiring no additional measurements by processing available plant data. It is also capable of estimating the efficiency of a chemical cleaning after restarting the plant and checking the evolution and kinetics of eventual TSP re-blockage.

Commentary by Dr. Valentin Fuster
2010;():157-162. doi:10.1115/ICONE18-29465.

Providing a reliable upper limit of radiological consequences to the plant personnel and the general public is typically the aim of a safety evaluation for anticipated operational occurrences or design basis accidents, as presented in a safety analysis report. A typical tool for dispersion calculation and dose evaluation is MACCS2. In the present analysis four types of calculations are presented: a first calculation, typical for licensing analysis, with the MACCS2 computer code. In a second step conservative assumptions e.g. ground release even if a stack release would be realistic, are dropped. In a third step calculation two is repeated with RODOS, a code (online decision making tool) used to predict the radiological consequences of an accidental release of activity. The step three calculation still contains all the conservative assumptions that are built in the MACCS2 code. In a last step these assumptions are removed, and a “best estimate” calculation on the dose to the public is performed. The whole analysis (step one to four) is repeated for different source terms (noble gases only, tritium dominated, primary system water [[ellipsis]]) and for different weather conditions. Two main conclusions can be drawn. The first by comparing step two (MACCS2) and step three (RODOS). Here the boundary conditions of the calculations are set to be as similar to each other as possible. The paper shows that despite the fact that MACCS2 uses a Gaussian plume model, while RODOS uses a puff model for dispersion calculation, doses of the same order of magnitude are calculated. For the second conclusion the step one (MACCS2, conservative) and step four (RODOS, best estimate) calculations are compared, it is shown that although the margin of conservatism varies considerably from case to case, the results differ at least one order of magnitude.

Topics: Safety , Licensing
Commentary by Dr. Valentin Fuster
2010;():163-169. doi:10.1115/ICONE18-29470.

There are an increasing number of nuclear reactors around the world operating well beyond their original design lives. Some pump seals supplied by original equipment manufacturers are either no longer available or their performance does not meet current requirements. This paper describes both the testing and operational experience of two nuclear pump seals that have been developed by AECL to replace the original equipment manufacturer’s seals in AECL’s National Research Universal reactor, which has been operating since 1957. The two seals described are the main heavy water pump seal and the heavy water degassing pump seal. The main heavy water pump seal is a tandem seal containing two identical, but 180° offset, eccentric seals of 95 mm balance diameter. The requirements were low pressure (0.3 MPa), low leakage (< 0.2 mL/min) and long, reliable lifetime (seven years). The seals were arranged to put full pressure across the outboard seal initially, with the inboard seal providing complete back up in the event of an outboard seal failure. Laboratory testing and operational performance, since initial installation in 1993, is described. The heavy water de-gassing pump seal design is based upon the main heavy water pump seal (a tandem seal containing two identical, but 180° offset, eccentric seals) but its balance diameter of 65 mm is much smaller than that of the main heavy water pump seal. Laboratory testing and operational performance, since initial installation in 2001, is also described.

Commentary by Dr. Valentin Fuster
2010;():171-176. doi:10.1115/ICONE18-29471.

The Steam Generator Asset Management Program (SGAMP) is a long term program designed to maximize the performance and reliability of the steam generators. The SGAMP focuses on plant specific conditions and hence is applicable to the original or the replacement steam generators. It is recommended that the utility and the vendor form a joint steam generator management team (SGMT) to develop, monitor and implement a long-term plan to address steam generator operation, maintenance and life extension goals. The SGMT will consist of representatives from operations, chemistry, maintenance and engineering functions and will be responsible for making decisions related to the steam generators. The charter of the SGMT is to develop a steam generator strategic plan that will cost-effectively manage steam generator options. The strategic plan is consistent with the Steam Generator Program Guidelines (NEI 97-06 in the United States). The strategic plan is a living document and is revised periodically to incorporate inspection results, new technology developments, lessons learned and industry experience. Cost-benefit analyses of strategies may be performed to prolong steam generator operability through steam generator performance modeling (tube degradation, fouling, etc.), diagnostic tools, regulatory strategy, condition monitoring and operational assessment strategy, and maintenance strategy. The SGMT will provide input regarding potential maintenance of the steam generators with schedule and cost impacts for each outage. It will also recommend engineering evaluations to be performed in support of program goals and will develop short- and long-term recommendations. These recommendations will address action plans, performance measures and results. Secondary side inspection and cleaning strategy should be developed (techniques and frequency) to maximize performance cost-effectively. This paper is based on Westinghouse experience gained by working with several pressurized water reactor (PWR) plant operators in the United States (US).

Topics: Boilers
Commentary by Dr. Valentin Fuster
2010;():177-182. doi:10.1115/ICONE18-29578.

We confirmed defect detection performance of a remote field eddy current testing (RFECT) in order to inspect a helical-coil-type double wall tube steam generator (DWTSG) with a wire mesh layer for the new-type small fast reactor 4S (Supersafe, Small and Simple). As high sensitivity techniques, to increase an indirect magnetic field intensity, we focused attention on increasing a direct magnetic field intensity in the vicinity of an exciter coil by the use of an exciter coil with a magnetic material (flux guide). We adopted horizontal type multiple detector coils with flux guides arrayed circumferentially to enhance sensitivity of radial direction. According to the experimental results, the indirect magnetic field intensity (the voltage of detector coil in the region of indirect magnetic field) increased more than 100 times by the application of the exciter and detector coils with flux guides. Finally, we detected the pinhole defect of 1 mm in diameter and 20% of outer tube thickness in depth over the wire mesh layer by the adoption of the flux guides and horizontal type multiple detector coils. And we confirmed that the RFECT probe is also useful to detect the thinning defects.

Commentary by Dr. Valentin Fuster
2010;():183-186. doi:10.1115/ICONE18-29591.

With the improvement of nuclear technology and nuclear safety, accidental operating procedure, as an important component of the defence in depth concept for nuclear power plant operation, develops in the way of being more robust and more operator-friendly, especially following the Three Miles Island accident. In China, nuclear power plants try to change from Event Oriented Procedure (EOP) to State Oriented Procedure (SOP) which is a scenario independent approach, developed by French institutes. This paper tries to discuss the principle of the state oriented approach and the composition of SOP. At the same time, comparison between SOP and EOP is made. Having understood the bases of SOP, China Nuclear Power Plants apply it from design to implementation appropriately only.

Commentary by Dr. Valentin Fuster
2010;():187-194. doi:10.1115/ICONE18-29608.

While risk-informed in-service inspection (RI-ISI) program has been applied in several countries to enhance the traditional periodic inspection program (PIP), many other countries are waiting for more successful implementation experiences to be accumulated. Canadian Nuclear Safety Commission (CNSC), the regulatory body of nuclear industry in Canada, became increasingly interested in the risk-informed decision making methodology. Several small-scale pilot studies on RI-ISI have been initiated by Canadian utilities during the past few years. Nevertheless, a RI-ISI methodology appropriate for the CANDU technology that can be accepted by the stakeholders has yet to be developed. The development of the RI-ISI methodologies derived from the PWR/BWR operating experiences is first reviewed, followed by an examination of Canadian periodic inspection standard CSA N285.4 and its evolution from a RI-ISI perspective. Finally several key technical issues and research needs in developing an advanced RI-ISI methodology for nuclear power plants are identified.

Commentary by Dr. Valentin Fuster
2010;():195-200. doi:10.1115/ICONE18-29615.

The major threat that nuclear power plants (NPPs) pose to the safety of the public comes from the large amount radioactive material released during design-basis accidents (DBAs). Additionally, many aspects of Control Room Habitability, Environmental Reports, Facility Siting and Operation derive from the design analyses that incorporated the earlier accident source term and radiological consequence of NPPs. Depending on current applications, majority of Chinese NPPs adopt the method of TID-14844, which uses the whole body and thyroid dose criteria. However, alternative Source Term (AST) are commonly used in AP1000 and some LWRs (such as Beaver Valley Power Station, Units No. 1 and No. 2, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2, Kewaunee Power Station and so on), so it is attempted to adopt AST in radiological consequence analysis of other nuclear power plants. By introducing and implementing the method of AST defined in RG 1.183 and using integral safety analysis code, a pressurized water reactor (PWR) of 900 MW nuclear power plant analysis model is constructed and the radiological consequence induced by Main Steam Line Break (MSLB) accident is evaluated. For DBA MSLB, the fractions of core inventory are assumed to be in the gap for various radionuclides and then the release from the fuel gap is assumed to occur instantaneously with the onset of assumed damage. According to the assumptions for evaluating the radiological consequences of PWR MSLB, dose calculation methodology is performed with total effective dose equivalent (TEDE) which is the criteria of dose evaluation. Compared with dose criteria of RG 1.183, the dose of control room, exclusion area boundary and outer boundary of low population zone are acceptable.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2010;():201-206. doi:10.1115/ICONE18-29656.

With the research and development of nuclear power conception engineering and the requirement of structural integrity & durability, mass concrete structure was adopted more frequently in nuclear power construction project engineering. It caused that the construction requirements were more strictly on how to ensure construction quality of mass concrete for the nuclear power project which related to the nuclear safety at the first AP1000 plants in the world. Facing this challenge, exploring and practicing activities were taken on how to ensure mass concrete construction quality for the first AP1000 unit at Sanmen site with the considering of the quantity of concrete more than 2000 cubic meters with single pour seldom applied on nuclear power project. Based on the analysis and calculation factors which influence the mass concrete quality, the measurement and methodology system of mass concrete quality assurance was developed through the path of administration and technical. Success experience obtained and the preset quality realized through several mass concrete project construction activities on site. This thesis mainly introduce the detailed measurements explored and practiced for ensuring mass concrete construction quality at Sanmen nuclear power project construction site. The success experiences were generalized and summarized.

Commentary by Dr. Valentin Fuster
2010;():207-212. doi:10.1115/ICONE18-29700.

A maintenance strategy to repair or modify the dissimilar weld on the branch connection of the steam generator hot leg in CANDU6 was developed. The repair method is to add a carbon steel pipe spool as a transition fitting to transfer the dissimilar weld to shop welding, and the modification method is to use the backup tap fabricated during plant installation, so that the weld quality would be superior comparing with field welding, and the quantity of argon gas getting into the primary heat transport system would be reduced simultaneously. A gasbag is used to isolate the branch connection for welding performance when the primary heat transport system is drained to low level state during the prolonged shutdown. If the gasbag could not work, a nonstandard pipe fitting with O-ring should be used to isolate the branch connection.

Commentary by Dr. Valentin Fuster
2010;():213-221. doi:10.1115/ICONE18-29701.

An appropriate flange tightening methods for small bore and low rating piping flange joints are clarified to improve the sealing performance of the bolted flange joints and the workability of flange tightening work. It is said that lubricant on the screw of the bolts and the nut-seating surface can minimize the variability of axial force acting on flange bolts, while this process might make it harder to tight the bolts uniformly especially for small bore low rating flanges. So, in this paper the appropriate condition to apply lubricant is clarified by a series of bolt tightening tests and sealing tests results. On the other hand, for the bolted flanges applying spiral wound gaskets, measuring the gasket compress dimensions help us to prevent uneven tightening balance and to perform the appropriate tightening work. Appropriate gasket compress dimensions are also clarified to ensure the sealing performance for the flanges based on the sealing tests results. Based on these test results, recommended flange-tightening methods have been summarized as an instruction and tightening work procedure to improve the sealing performance of the bolted flanges and the workability of flange tightening work.

Commentary by Dr. Valentin Fuster
2010;():223-228. doi:10.1115/ICONE18-29702.

Digital radiography is getting one of the common radiographic testing techniques in various industries now. However, to apply this new technique to nuclear components radiographic testing, one big issue is how we can evaluate and ensure that the taken images have enough image qualities to be used as inspection record. In film radiography, the IQI, which stands for Image Quality Indicator, have been used to ensure that taken films have enough quality to detect any specified defects in the products. So in this paper, new alternative IQI that developed in our previous study for digital radiography to evaluate digital image quality are tested and evaluated. In addition, image evaluation criteria are also developed and evaluated by calculating MTF, which stands for Modulation Transfer Function, from the IQI images taken with the products. Finally, the recommended procedures to evaluate radiographic testing image are summarized.

Commentary by Dr. Valentin Fuster
2010;():229-234. doi:10.1115/ICONE18-29703.

The excessive maintenance of the nuclear power plants (NPPs) may cause the early (infant) failure in Japan. An easy analysis; the Weibull analysis was applied to the evaluation of the failure mode. The Weibull analysis needs the hazard data. The maintenance information of the equipment which caused plant shutdown was required for the hazard calculation. However, maintenance information of the equipment was not open. Therefore, all equipment was assumed to be maintained during every shutdown. This assumption was based on renewal process. However, a repair after unplanned shutdown of NPP is generally a restoration of only failed function without system overhaul. The system must be considered to age continuously. The system was not renewed. The operation data must be regarded as one continuous data before and after unplanned shutdown. An improvement of the Weibull analysis was required for NPPs. The model of the Weibull analysis was investigated. The competitive model in which shutdown caused by other than focused equipment/cause may be supposed to be continuous data could not be applied for a comprehensive analysis. Furthermore, the calculation method of the Weibull analysis was investigated. The calculation method of the hazard was viewed. A denominator of the hazard is the number of data which is cut for every continuous data by renewal process. However, multiple considerations of operation periods before unplanned shutdowns might cause underestimation of the failure rate in case of restoration process. Therefore, a dominator of the hazard was not supposed to be the number of data but the number of survived equipments (plants) at each time according to the definition of the hazard. This improved method is for the restoration process. The performance of Japanese NPPs was evaluated by improved method. The failure modes of Japanese NPPs were early failure modes. Moreover, performances of U.S. NPPs was tried to be evaluated by improved method. Operation data was collected from “NRC Power Reactor Status Reports”. However, many “maintenance outage”s which are the shutdowns of unknown origin were found. Therefore, DOE information was supplemented to investigate the “maintenance outage”. Failure modes of U.S. NPPs were the early failure modes, and failure rates were larger than Japanese NPPs.

Commentary by Dr. Valentin Fuster
2010;():235-242. doi:10.1115/ICONE18-29733.

Exelon Nuclear has recently installed adjustable speed drives (ASD), or variable frequency drives (VFD), in place of the original motor-generator (MG) sets for their boiling water reactor (BWR) recirculation pumps at Quad Cities. This paper reviews expected versus actual performance of the drives and motors. The discussion focuses on energy savings, motor starting characteristics, control accuracy and stability, motor and cable thermal behavior; as well as, a comparison of actual supply system input harmonic measurements versus analysis results. Included in the review are a few operational lessons learned with regard to the startup process, input medium voltage low speed holds, loss of cooling water pump suction pressure, cooling system surge tank level indication, and other miscellaneous points of interest.

Commentary by Dr. Valentin Fuster
2010;():243-250. doi:10.1115/ICONE18-29738.

This paper indicates the importance of classified management of components in view of different functions of plant components, presents the principles based on which Third Qinshan Nuclear Power Plant (TQNPP) implements the classified management of components, and introduces the concept of SPV component in nuclear power stations. It focus on expounding the analysis and identification of SPV systems and components, explaining the methods to determine the list of SPV key systems and to evaluate the system priority sequence; getting the SPV fault tree of the system and SPV points; classification for the SPV points, finding the SPV component chain. The management requirements and practice for SPV components are discussed from different viewpoints of component management.

Commentary by Dr. Valentin Fuster
2010;():251-259. doi:10.1115/ICONE18-29746.

Qinshan Phase III is the first commercial pressurized heavy water reactor (PHWR) NPP in China, and it uses CANDU-6 design developed by AECL. Based on plant design and operation experience, the event tree analysis model has been developed for both small break LOCA (SB-LOCA) and large break LOCA (LB-LOCA), which is an important aspect of operational Probabilistic Safety Assessment (PSA). Both SB-LOCA and LB-LOCA event tree analysis have been performed for Qinshan Phase III CANDU-6 PHWR NPP (TQNPC). And the event sequence development and plant damage status (PDS) were provided in the analysis. It reflects actual plant configuration and response under a certain event, and various break type and locations were also considered in the event tree analysis, e.g. Pressure Tube Rupture, Pressure Tube and Calandria Tube Rupture, Feeder Breaks, Pressurizer Relief/Steam Bleed Valves Fail Open, Liquid Relief Valves Fail Open, etc.

Commentary by Dr. Valentin Fuster
2010;():261-265. doi:10.1115/ICONE18-29802.

Optimized and improved measures have been implemented in Qinshan III NPP to optimize the management of routine production plan, strengthen maintenance work risk analysis, and improve the plan execution capability, which involve unified management of generation, refuelling, periodic test and maintenance plans, simplifying the defect scales and reducing interlinks of defect disposal, intensifying the assessment on plan execution and adopting performance evaluation and star rating measures.

Commentary by Dr. Valentin Fuster
2010;():267-271. doi:10.1115/ICONE18-29803.

Severe weather such as typhoon has long been a great challenge threats the safe operation of nuclear power plants. To cope with typhoon, Qinshan III NPP has developed an effective management system, including building powerful organizations, creating standard response procedures and consumable storage, which proven to be effective to ensure the safe operation of Qinshan III plant under severe weather conditions.

Commentary by Dr. Valentin Fuster
2010;():273-277. doi:10.1115/ICONE18-29812.

The functional hierarchy and the performance degradation characteristic of the canned motor pump (CMP) used in nuclear power plants (NPP) are analyzed. And to model the performance degradation of CMP, a hierarchical model based on multi-Agent system (MAS) is proposed. In the model, the framework of the system is designed and the disadvantages of current assessment are conquered. According to the function structure of CMP, the performance degradation and prediction Agents for each hierarchy are divided and the concrete functions for each Agent are also defined. The MAS model provides a guidance template to implement equipment performance degradation and prediction system.

Commentary by Dr. Valentin Fuster
2010;():279-287. doi:10.1115/ICONE18-29826.

In this article, the effects of the non-propagating open cracks on the dynamic behaviors of a cantilevered pipe conveying fluid are studied. The model divides the pipe into a number of segments from the crack sections and assembles all segments each by each by a rotational spring which has no mass. The stiffness of the spring is obtained through linear fracture mechanics. In order to obtain the modal functions which satisfy the boundary conditions and geometrical discontinuity conditions at the crack’s location, a simple approach is used. That is adding polynomial functions to the modal functions of the uncracked beam. The equations of motion for the cracked cantilevered pipe conveying fluid is derived based on the extended Lagrange equations for systems containing non-material volumes. Not only the virtual work done by the discharged fluid, but also that done by the fluid at the crack position due to the geometrical discontinuity conditions are considered in the present equations of motion. In this article, several numerical examples are given. The comparisons of solutions of the present equations with that of model in existence show that the present work is better. The influences of the relative depth, the position ratio of the cracks, the flow velocity on the eigenvalues are depicted.

Topics: Fluids , Pipes
Commentary by Dr. Valentin Fuster
2010;():289-294. doi:10.1115/ICONE18-29843.

The cracks detected by in-service inspection are not always removed when they are not hazardous according to fitness-for-service evaluations. In such cases, it is important to monitor the growth of these cracks in order to assess the validity of their integrity assessment. However, due to the limitation of the accuracy of size determination by ultrasonic testing, it is difficult to know how much the cracks have grown since the previous measurement. In this study, a method for crack growth monitoring during plant operation is proposed. When a pipe is deformed elastically due to internal pressure, the strain at its external surface increases. By measuring the change in strain for the outside of the cracked pipe continuously, it is possible to determine how much the crack size changes. Elastic finite element analyses were performed for cracked pipes under internal pressure. From the analyses under various crack sizes, the accuracy and resolution of the proposed monitoring method were evaluated. It was revealed that the method could detect crack growth of less than several hundred micrometers.

Commentary by Dr. Valentin Fuster
2010;():295-299. doi:10.1115/ICONE18-29845.

Reactor thermal power uprate (Power uprate) of operating light water reactors has long successful experiences in many nuclear power plants in the United States of America and European countries since late 1970’s. And it will be also introduced in Japan soon. This paper mainly describes the outline of the attempt of five-percent reactor thermal power uprate of Tokai No.2 Nuclear Power Station (Tokai-2) operated by the Japan Atomic Power Company (JAPC). It will be the leading case in Japan. Tokai-2 is GE type Boiling Water Reactor (BWR) of 1100 MW licensed electric power output and it commenced commercial operation in November 28, 1978. Power uprate is an effective approach for increasing electric power output. And it is recognized as one of the measures for effective and efficient use of existing Japanese operating nuclear power plants. It can contribute to inexpensive and stable electric power supply increase. Especially “Stretch Power Uprate (SPU)” requires only minor equipment modification or component replacement. It is also a countermeasure against global warming. Therefore it is a common theme to be accomplished in the near future for both Japanese electric power companies and government. JAPC started feasibility studies on power uprate in 2003. And in 2007, JAPC established a plan to achieve five-percent power uprate in Tokai-2 and announced this project to the public. This is a leading attempt in the Japanese electric power companies and it is the first case under the current Japanese regulatory requirements. In this plan, JAPC reflected lessons learned from preceding nuclear power plants in the United States and European countries, and tried to make most use of the performance of existing systems and components in Tokai-2 which have been periodically or timely renewed by utilizing more reliable and efficient design. JAPC plans to submit application documents to amend current License for Reactor Establishment Permit shortly. It will contain a complete set of revised safety analysis results based on the uprated reactor thermal power condition. Successful introduction of Tokai-2 power uprate will contribute to the establishment of regulatory process for power uprate in Japan and following attempts by other Japanese electric power companies.

Topics: Power uprate
Commentary by Dr. Valentin Fuster
2010;():301-304. doi:10.1115/ICONE18-29894.

Dr. Mainte, an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) has been developed based on PFM (Probabilistic Fracture Mechanics) analyses. The concept of the simulator is to provide a decision-making system to optimize maintenance activities for representative components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability and economic efficiency.

Commentary by Dr. Valentin Fuster
2010;():305-309. doi:10.1115/ICONE18-30033.

Two ultrasonic testing (UT) devices to inspect the internals of nuclear reactor have been developed. The one is a jet pump UT device to inspect the inner weld line of the jet pump, and the other is a shroud UT device to inspect the outer weld line of the shroud. The jet pump UT device is mainly composed of an inspection probe scanner and a wheeled platform with a telescopic guide. Since the inspection probe scanner has been designed slim enough to pass through the narrow opening of the jet pump nozzle, it can be remotely positioned inside the jet pump transported by the wheeled platform. The shroud UT device is mainly composed of a flat-type remotely operated vehicle (ROV) and a positioning mast of ROV. The ROV is installed remotely on the outer surface of the shroud using the positioning mast. And the ROV has been designed thin enough to pass through the narrow gap between the jet pump and the shroud, so that it can move horizontally on the surface of the shroud with automatic cable feeding. Consequently, the proposed remote and automatic inspection devices can perform the inspection work in short time without using fuel handling machine (FHM). Therefore, the inspections can be performed simultaneously with the refueling work, which contributes to the shortening of regular inspection periods of nuclear power plants.

Commentary by Dr. Valentin Fuster
2010;():311-318. doi:10.1115/ICONE18-30057.

The purpose of this work is to test the capability of TRACE5 code in the simulation of thermal-hydraulic transients concerning Condensation-Induced Water Hammer (CIWH) phenomena in a horizontal branch pipe connected to the vessel downcomer. The CIWH is produced by the condensation of the steam by subcooled water counterflow in the horizontal pipe, which causes two-phase flow interfacial instability, and is capable of initiating a severe water hammer, possibly leading to significant plant damage. The work is developed in the frame of OECD/NEA ROSA Project Test 2, performed in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA) [1]. The purpose is to provide an analytical model for the LSTF installation, in order to evaluate the critical inlet water flow rates and system pressures of CIWH in a long horizontal pipe without using interfacial friction factor or heat transfer coefficients and using the default TRACE criterion of transition from stratified to a slug flow. The analytical model is designed with the thermal-hydraulic code TRACE5 via 1D-components, reproducing the actual branch where the CIWH is produced. A TEE component is connected to a FILL component, which simulates the water injection, and to a BREAK component set to the boundary conditions that simulate the downcomer. Our model uses one-dimensional flow equations and default correlations of interfacial shear stress and heat and mass transfer available by TRACE. Several comparisons are performed, varying pressure system and water injection mass flow rates. Simulated pressure pulses are characterized, studying parameters such as geometry nodalization, time-step effect, Courant number, numerical diffusion, etc. Results show that 1D model slightly underestimates the maximum pressure pulse intensity in all cases considered.

Commentary by Dr. Valentin Fuster
2010;():319-323. doi:10.1115/ICONE18-30150.

Nuclear power plants commissioning is an entire inspection and verification process of systems and equipments carried out by commissioning party, to prove that the overall performance of system, equipment and unit can meet the design requirements, operation criteria and nuclear safety criteria, and the plant is qualified for long-term stable and safe operation. As commissioning work has the characteristic of long lasting time, broad scope, high technology and lots of interfaces, and so on, it must be carried out under a set of scientific management system, to accord with requirements of “six controls” of commissioning work. The text aims at researching and investigating into commissioning management system that is in accordance with CPR1000 nuclear power plant commissioning through analysis of constitutes and practice situation of Ling AO phase II project commissioning management system, so as to instruct and standardize the commissioning management of nuclear power plants.

Commentary by Dr. Valentin Fuster
2010;():325-334. doi:10.1115/ICONE18-30238.

Visual observation of inner side of a reactor pressure vessel of Japan Materials Testing Reactor (JMTR) was carried out using an underwater camera before the JMTR refurbishment work from the view point of its long term utilization, because the reactor pressure vessel of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed using the gauze were observed around nozzles in a top closure of the reactor pressure vessel. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the reactor pressure vessel was observed, and then the integrity of the reactor pressure vessel was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the reactor pressure vessel of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the reactor pressure vessel in future.

Topics: Reactor vessels
Commentary by Dr. Valentin Fuster
2010;():335-343. doi:10.1115/ICONE18-30240.

The recent NRC Generic Letter (GL) 2008-01 titled “Managing Gas Intrusion in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems” to nuclear power plant licensees in the United States requires demonstration of suitable design, operational testing and control measures in order to maintain licensing commitments [1]. The generic letter outlines a number of actions that are detailed in nature; such as establishing pump void tolerance limits, limits on pump suction void fractions, etc. Each addressee was requested to evaluate their Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) system, and Containment Spray (CS) system. For each of these systems, design, operation, and test procedures were evaluated to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 Code of Federal Regulation (CFR) 50 Appendix B [2]. In the GL 2008-01, licensees were requested to evaluate the ECCS, DHR and CS systems along four principal areas to ensure that gas accumulation is maintained less than the amount that challenges the operability of the systems, and that licensees shall take appropriate actions when the conditions are identified. The four principal areas are licensing basis, design, testing, and corrective actions. Each addressee was requested to provide a summary description of how the “REQUESTED ACTIONS” in the generic letter were addressed within nine months of the generic letter issue date. If an addressee determined that system or procedure modifications were necessary based on the review of the requested actions but cannot be accomplished within nine months of the date of the generic letter, then the addressee should provide a plan and schedule for completion of the actions. Many plants used their corrective action programs to accomplish this task. In its response, the licensee addressed any alternative course of action that it proposed to take, including the basis for the acceptability of the proposed alternative course of action. The nuclear industry, under the auspices of the Nuclear Energy Institute, has worked collaboratively with the industry to develop solutions and responses to the nine month NRC request, and these responses were submitted in October of 2008. Since that time, the NRC has been reviewing the plant submittals and issuing requests for additional information (RAIs) to the plants for clarification of their respective programs. This paper provides a snapshot review of the regulation of gas voids in the United States by focusing not only on industry actions to address the generic letter but also on the nature of the NRC requests to the nuclear plants for clarification of plant gas mitigation programs. The goal of the paper is to explore if the RAIs will provide some insights on NRC expectations of the industry as plants address gas intrusion in safety related Nuclear Steam Supply Systems (NSSS).

Topics: Steam
Commentary by Dr. Valentin Fuster
2010;():345-354. doi:10.1115/ICONE18-30261.

Two-phase gravity-driven drainage systems are used in many applications within nuclear power Balance of Plant (BOP) applications such as the drain lines for moisture separator re-heaters (MSRs) and feedwater heaters. Design of these systems is typically based on industry-oriented guidelines and operator-based experience. Changes in plant operation, such as uprates and equipment modification and/or replacement, are relatively common as plants seek to generate more power with greater efficiency. These plant modifications may inadvertently change system operation from design conditions and impose undesirable system transients. This paper seeks to provide a method for analyzing BOP drainage systems in an effort to characterize and mitigate drain flow transients. Previous methodologies diagnose and evaluate drain instability through measurement, empirical analysis, and operational experience. This paper identifies methods that can be utilized to generate computational models of discrete plant drainage systems that decrease the level of speculation involved in previous analyses. Additionally, a real-world application of this method is presented to demonstrate how computer modeling can accurately mimic plant transients.

Commentary by Dr. Valentin Fuster
2010;():355-360. doi:10.1115/ICONE18-30281.

It is well-known that RCM is an advanced and effective maintenance strategy in practice. With the development of the automation and mechanization in modern industry, RCM method turns to be complex and consumes more resources in real production. However, the development and application of the Streamline RCM (SRCM) has injected new vitality for the new situation, especially in the nuclear power plants. This paper firstly introduces the background, the characteristics of the SRCM and the differences from RCM, and then shows the process in detail as well as the application status of the SRCM in country and abroad. It is proved that SRCM is a unique available method which saves the time and resource consumed, ensuring the integrity and correctness of the classical RCM. Finally, the weak points and the prospect are reviewed and prospected.

Commentary by Dr. Valentin Fuster
2010;():361-365. doi:10.1115/ICONE18-30291.

The article presents how to plan, organize, and put in practice of the 18-months fuel cycle project on the newly built up nuclear plants. Some problems related to the field experience of modification and the amelioration suggestions are also provided. It’s involved that operation procedure modification, operation technical specification modification, test period prolongation, set point adjustment, boric acid concentration adjustment, re-start physical experiment and so on.

Commentary by Dr. Valentin Fuster
2010;():367-376. doi:10.1115/ICONE18-30329.

Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes. This may imply the use of a dedicated fuel rod thermo-mechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermo-mechanical code. The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes (RELAP5-3D©) calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, with the aim of a better understanding of the uncertainties involved and their technological consequences on the behavior of the fuel rods, not addressed in the current paper.

Topics: Failure , Fuel rods
Commentary by Dr. Valentin Fuster
2010;():377-381. doi:10.1115/ICONE18-30355.

In Korea, newly advanced installation method of containment liner plates was developed and adopted for the construction of Shin Wolsong Unit 2 NPP. The key method of the new installation method of concrete liner plates was using a three-stage-based modularization (lifting of 3 plate ring module) instead of using conventional installation method. In this paper, overall newly developed installation method of containment liner plates including such as characteristics, merits and procedures were described briefly.

Commentary by Dr. Valentin Fuster

Fuel Cycle and Decommissioning

2010;():383-386. doi:10.1115/ICONE18-29006.

The sorption of 131 I− , 131IO3, I2 and CH3131I from water solutions at 25°C on new composite materials obtained by modifying of cation-exchange resin KU-2 was investigated. It was established, that the given materials are capable to absorb I2 both from distilled water, and from a water coolant of the WWER-type NPPs, with distribution factors Kd more than 103 cm3 /g at V/m = 100. Thus, it was found, that practically full I2 absorption (more than 95.0%) was achieved for 15 min. It was shown, that anion-exchange resin AV-18 is capable to adsorb CH3131I from a water solutions with distribution factors Kd more than 300 cm3 /g at V/m = 100.

Commentary by Dr. Valentin Fuster
2010;():387-391. doi:10.1115/ICONE18-29028.

Acetohydroxamic acid (AHA) is an organic ligand planned for use in the Uranium Extraction (UREX) process. It reduces neptunium and plutonium, and the resultant hydrophilic complexes are separated from uranium by extraction with tributyl phosphate (TBP) in a hydrocarbon diluent. AHA undergoes hydrolysis to acetic acid which will impede the recycling of nitric acid. During recent discussions of the UREX process, it has been proposed to replace AHA by formohydroxamic acid (FHA). FHA will undergo hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. The reported reduction potentials of AHA and pertechnetate (TcO4) indicated that it may be possible for AHA to reduce technetium, altering its fate in the fuel cycle. At UNLV, it has been demonstrated that TcO4 undergoes reductive nitrosylation by AHA under a variety of conditions. The resulting divalent technetium is complexed by AHA to form the pseudo-octahedral trans-aquonitrosyl (diacetohydroxamic)-technetium(II) complex ([TcII (NO)(AHA)2 H2 O]+ ). In this paper, we are reporting the synthesis of FHA and its complex formation with technetium along with the characterization of FHA crystals achieved by NMR and IR spectroscopy. Two experiments were conducted to investigate the complexation of FHA with Tc and the results were compared with previous data on AHA. The first experiment involved the elution of Tc from a Reillex HP anion exchange resin, and the second one monitored the complexation of technetium with FHA by UV-visible spectrophotometry.

Commentary by Dr. Valentin Fuster
2010;():393-397. doi:10.1115/ICONE18-29057.

The two Units of TianWan Nuclear Power Station (phase 1) are in their third fuel cycle operation now. China Nuclear Power Engineering Co., Ltd. (CNPE) has finished the reload design of Unit 1 and the check on Russia’s computation of Unit 2 in the third cycle. From the forth cycle on, the in-core reload design will be given by Chinese part. It is summarized in this paper the problems met and the way to solve them during our design process. We discussed how to retain the related parameters, such as maximum burn-up and power peak factor, in reasonable ranges in the first part. In the last part, four design cases which can be used in cycle 5 for Unit 1 are recommended and evaluated briefly. The first two cases are partly low-leakage schemas; they satisfy the needs of the related restrictions very well. Case 3 is completely a low-leakage schema and case 4 is a long life schema. For case 3 and 4, a few new kind of assemblies, which are different from the present assemblies in uranium enrichment or the numbers of Oxide Gadolinium fuel rods, may be adapt to for better parameters.

Commentary by Dr. Valentin Fuster
2010;():399-403. doi:10.1115/ICONE18-29058.

With the sustained and rapid development of the nuclear power plants, the spent fuel which is produced by the nuclear power plants will be rapidly rising. Spent fuel is High-level radioactive waste and should be disposed safely, which is important for the environment of land, public safety and health of the nuclear industry, the major issues of sustainable development and it is also necessary part for the nuclear industry activities. It is important to study and resolve the high-level radioactive waste repository problem. Spent nuclear fuel is an important component in the radioactive waste, The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The ductile cast iron insert provides the mechanical strength whereas the copper protects the canister from corrosion. The canister inserts material were referred to as I24, I25 and I26, Spent nuclear fuel make the repository in high radiant intensity. The radiation analysis of canister insert is important in canister transport, the dose analysis of repository and groundwater radiolysis. Groundwater radiolysis, which produces oxidants (H2 O2 and O2 ), will break the deep repository for spent nuclear fuel. The dose distribution of canister surface with different kinds of canister inserts (I24, I25 and I26) is calculated by MCNP (Ref. 1). Analysing the calculation results, we offer a reference for selecting canister inserts material.

Commentary by Dr. Valentin Fuster
2010;():405-411. doi:10.1115/ICONE18-29066.

Acidization was studied on a uranium sandstone deposit in Inner Mongolia with low–permeability and heavy calcium cementation. Acid dissolving test indicates that hydrochloric acid, formic acid and mud acid can easily dissolve formation minerals. With proper volumes and concentrations of acids used, the risk of precipitation of reaction products could be minimized. Core flow acidizing trial shows that the acidic fluid systems of hydrochloride acid, formic acid or acetic acid can significantly improve the core permeability. The average permeability has increased by 763 percent for the above three systems. But mud acid didn’t increase the core permeability; on the contrary, it caused formation damage, and led to lowering permeability. In the pilot test, the injection rate has improved by 118 percent for 6 wells. The acid treatment results indicate that a significant production enhancement of wellfields can be achieved by acid stimulation.

Commentary by Dr. Valentin Fuster
2010;():413-417. doi:10.1115/ICONE18-29075.

The radiological characterization includes: the collection and review of historical file; the performing calculation of radionuclide inventory in the reactor; in situ measurement; sampling analyses; the review and evaluation of the data obtained; the comparison of calculated result with measured date etc. The special attention should put to the information from the key part in reactor for end state radiological characterization. The sampling from the “hot spot” should not be lost; the number of the sampling should reasonable base on reliable statistics. The radioactivity density for site release should comply with the guide, standard and regulation of international atomic energy agency and our country.

Commentary by Dr. Valentin Fuster
2010;():419-427. doi:10.1115/ICONE18-29091.

A fine crystalline ammonium tungstophosphate (AWP) exchanger with high selectivity toward Cs+ was encapsulated in biopolymer matrices (calcium alginate, CaALG). The characterization of the AWP-CaALG microcapsule was examined using SEM/WDS, IR and DTA/TG analyses, and the selective separation and recovery of 137 Cs were examined by the batch and column methods using simulated and real high-level liquid waste (HLLW). The free energy (ΔG0 ) of the ion exchange (NH4+ ↔ Cs+ ) for fine AWP crystals was determined at −13.2 kJ/mol, indicating the high selectivity of AWP towards Cs+ . Spherical and elastic AWP-CaALG microcapsules (∼700 μm in diameter) were obtained and fine AWP crystals were uniformly immobilized in alginate matrices. Relatively large Kd values of Cs+ above 105 cm3 /g were obtained in the presence of 10−3 ∼1 M Ca(NO3 )2 , resulting in a separation factor of Cs/Rb exceeding 102 . The irradiated samples (60 Co, 17.6 kGy) also exhibited large Kd values exceeding 105 cm3 /g in the presence of 2.5 M HNO3 . The Kd values in the presence of 0.1–9 M HNO3 for 67 elements were determined and the order of Kd value was Cs+ ≫ Rb+ > Ag+ . The breakthrough curve of Cs+ had an S-shaped profile, and the breakpoint increased with decreasing flow rate; the breakpoint and breakthrough capacity at a flow rate of 0.35 cm3 /min for the column (0.7 g AWP-CaALG) were estimated at 25.2 cm3 and 0.068 mmol/g, respectively. Good breakthrough and elution properties were retained even after thrice-repeated runs. The uptake (%) of Cs+ in simulated HLLW (28 metal components-1.92 M HNO3 , SW-11, JAEA) was estimated at 97%, and the distribution of Cs+ and Zr/Ru into the AWP and alginate phases, respectively, were observed by WDS analysis. Further, the selective uptake of 137 Cs exceeding 99% was confirmed by using real HLLW (FBR “JOYO”, JAEA). The AWP-CaALG microcapsules are thus effective for the selective separation and recovery of Cs+ from HLLWs.

Commentary by Dr. Valentin Fuster
2010;():429-436. doi:10.1115/ICONE18-29098.

Reactivity Initiated Accident (RIA) leads to an unwanted increase in fission rate and power in a region of the reactor core confined around the position of occurrence. The power excursion due to such events may cause fuel rods failures and a subsequent release of radioactive material into the primary coolant of reactor, in severe cases, this release could damage nearby fuel assemblies. In nuclear power plants, RIAs are due to control system faults, e. g. control elements ejection/insertion, or rapid changes in temperature or pressure of moderator. In Boiling Water Reactors (BWRs), the control rod drop accidents (RDAs) at cold zero power have been deeply investigated, in fact, notwithstanding they are less frequent in comparison with the control rod ejection event in PWRs, in this kind of plant these conditions are the most severe in case of a RIA occurrence. RDA transient, comprised in the design basis events considered in safety analysis, may cause rod failures especially at high burnup. To simulate a RIA, a peaked power pulse is applied to a pre-irradiated and re-instrumented rodlet aiming at investigating the most important phenomena that could lead to the rupture of cladding tubes. This paper is focused on the investigation of the TRANSURANUS fuel performance code capability to predict the thermomechanical state of rodlets subjected to RIA tests. To this purpose the FK-1 test, carried out at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute (JAERI), was simulated. This experiment is part of a set of 12 tests performed at the NSRR facility to study the performance under a reactivity initiated accident of BWR rodlets with burnup between 41 and 61 MWd/kgHM . In the FK-1 test, a STEP I BWR rodlet, previously irradiated in the Fukushima Daiichi Nuclear Power Station (Unit 3) operated by the Tokyo Electric Power COmpany (TEPCO) up to 45 MWd/kgHM , was subjected to a peak enthalpy insertion of 544 J/g. In this paper the code findings for the FK-1 test are discussed on the basis of the experimental data and the predictions of other stand-alone codes for transient analysis. The FK-1 predictions of FRAPTRAN (2001), FALCON (2003) and SCANAIR (ver. 3–2) are reported. The choice of fuel relocation model and important cladding properties (swelling, thermal expansion, thermal conductivity) was made relying on preliminary calculations whose results are also presented. Notwithstanding a satisfactory agreement between predictions and experimental data and a good agreement in the presented code-to-code comparison were envisaged, these results also emphasized the need to improve the models for FGR, heat transfer to plenum. Investigations are also required to ascertain possible contribution from fission gas to pelles thermal expansion. Ongoing modeling activity, performed at the ITU Joint Research Centre, is focused on a new model for FGR, ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile) is expected, in the near term, to give a contribution to refine the model for plenum gas temperature. These activities should improve the description of RIA transient and further investigations on NSRR tests will be performed with newly developed models. The work presented in this paper will be part of ENEA contribution in FUMEX III project leaded by the International Atomic Energy Agency (IAEA) and aimed at the improvement of fuel codes predictions at high burnup.

Commentary by Dr. Valentin Fuster
2010;():437-444. doi:10.1115/ICONE18-29300.

The HTR-10 is a pebble bed High-temperature gas-cooled reactor (HTGR) with a nominal thermal power of 10 MW. The on-line nondestructive burn-up (BU) measurement for the spherical fuel elements with TRISO coated fuel particles is an important issue in the fuel management of the HTR-10. The HTR-10 employs an HPGe gamma spectroscopy system to determine the BU values of the discharged fuel elements, and the accuracy of the BU measurement system is a crucial issue. The calibration requirement and the absolute calibration method are discussed and analyzed in this paper. In this work, the long-lived fission product 137 Cs is demonstrated as the best BU indicator in the HTR-10, whereas the inner-calibrating source method using the ratio of the amounts of 134 Cs to 137 Cs is not suitable. The BU value of a fuel element is proportional to the intensity of 137 Cs gamma-ray and independent on the irradiation history of the fuel element, except for a small correction term related to the radioactive decay of 137 Cs. The efficiency of the measurement system is calibrated by using a 137 Cs standard source embedded into the center of a graphite sphere with the same size and the same graphite material as the fuel element, which was located at the measurement position of the fuel sphere, to eliminate the possible systematic errors from the measurement system. The difference in geometry and self-absorption between the BU measurement and the calibration measurement is corrected by applying the MCNP simulation. The error sources of the calibration results are analyzed, in which the uncertainty of the source activity (∼5%) is found to dominate. The other major error sources include the fluctuation of the uranium loading in the fuel elements, the errors of the nuclear data, the counting errors of the detector, and the correction procedure for the detection efficiency. More than 1,900 fuel elements discharged from the HTR-10 have been measured by far, and the BU values of these elements spread in the range from 8.4±0.5 GWD/t to 20±1.0 GWD/t, where the relative errors of the burn-ups consist of the calibration errors about 5.2% and the counting errors around the order of 1%. The behavior of the measured burn-ups has been discussed qualitatively.

Commentary by Dr. Valentin Fuster
2010;():445-450. doi:10.1115/ICONE18-29358.

Water-soluble oxa-diamide ligand, N,N,N′ ,N′ -tetra-methyl-3-oxy-pentane-1,5-diamid (TMPDA) has been synthesized and purified. Its crystal structure, melting point, decomposition temperature, solubilities in aqueous phase and organic phase, distribution ratio between aqueous and organic phase, etc. are reported. The effect of TMPDA concentration in aqueous phase and HNO3 concentration in the equilibrium aqueous phase on the extraction efficiency of La(III), Ce(III), Pr(III), Nd(III), Zr(IV), Fe(III), Y(III), Mo(VI), Ru(III) and Pd(II) by 30% TRPO/kerosene have been studied. The results indicate that TMPDA dissolve well in aqueous phase but almost insoluble in kerosene or 30%TRPO/kerosene in the bi-phase system. It can effectively reduce the extraction of Ln(III), Y(III) and Zr(IV) into 30%TRPO/kerosene at a moderate acid system (0.24mol/L∼0.27mol/L HNO3 ). TMPDA is a promising stripping agent for Ln(III), Y(III) and Zr(IV) from loaded TRPO.

Topics: Zirconium
Commentary by Dr. Valentin Fuster
2010;():451-455. doi:10.1115/ICONE18-29436.

AREVA as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries and to accommodate foreseen EPR™ Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector together with TN International (a subsidiary of AREVA NC) decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and other local foreign Safety Authorities requirements. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: • Preferential flooding, • Fuel rod array expansion (so called “bird caging” effect), • Fuel sliding, • Neutron absorber penalty, • [[ellipsis]]. The French criticality code package CRISTAL is used to check several configurations reactivity and derived safety margins. The CRISTAL code package relies on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask, containing two fuel assemblies, is designed to maximize fuel isolation inside the cask and with neighboring ones even for large array configuration cases. Few and proven industrial products are used: • Stainless steel for the structural frame, • Balsa wood for impact limiters, • BORA® resin as neutron absorber. The cask is designed to handle mainly the EPR™ fuel assembly type and may be extended to other contents such as APWR fuel assembly type. After a brief presentation of the computer codes and the description of the shipping cask, the CRISTAL calculation results as well as the applied uncertainties will be discussed.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2010;():457-463. doi:10.1115/ICONE18-29574.

By proposing a credible response to the growing demand for energy while emitting no greenhouse gas, nuclear power will more than likely expand in the future, leading to increased quantities of used nuclear fuel to manage. Recycling the energy from this used fuel and an efficient waste management are key components for a sustainable development of nuclear energy. Current closed fuel cycle enables 25% in uranium savings, reduces the volume of waste by a factor 5 while its radiotoxicity is divided by a factor 10. Excellent track records of existing recycling plants participate to the competitiveness of MOX (Mixed Oxide) and ERU (Enriched Reprocessed Uranium) fuels compared to ENU (Enriched Natural Uranium) fuel. By offering a sound solution for nuclear waste management, recycling also contribute to favour nuclear acceptance. The advantages of closed fuel cycle have been demonstrated by the successful policy of recycling implemented in Europe for more than 30 years, with 35 reactors using MOX fuel with an excellent return of experience. The AREVA industrial recycling platform (La Hague and Melox plants) has treated more than 24,500 tons of used fuel and fabricated more than 1,600 tons of MOX fuel. With the new AREVA EPR™ reactor, recycling will take another step forward, enabling MOX fuel loading of up to 100%, thus offering optimized management options of recycled fuel, and giving more flexibility to its customers.

Commentary by Dr. Valentin Fuster
2010;():465-468. doi:10.1115/ICONE18-29632.

With an increased population and an increasing demand for power, nuclear power has attracted an increasing attention and mass nuclear power plant have been built in different countries in the past several decades. At present, about ten thousands ton spent fuels are discharged from nuclear power plant every year and the estimated capacity will approximately add up to 5×105 ton. Therefore, spent fuel reprocessing, by which the co-extraction and separation as well as purification of Uranium and Plutonium could be realized and ensure the recycle of uranium resources and the management of nuclear waste, is a vital step in nuclear fuel cycle including two major strategies, i.e. once-through cycle and closed fuel cycle. It is worth noting that the utilization of MOX fuel made by plutonium mixed with uranium has been successfully achieved in thermal reactor. Fortunately, the middle experiment plant of china spent fuel reprocessing has been being debugged and will be operated completely in future two years. Various reprocessing schemes have been proposed for the extraction of actinides from fission products and other elements presented in spent nuclear fuel. However, after numerous studies of alternate reprocessing methods and intensive searches for better solvents, the PUREX process remains the prime reprocessing method for spent nuclear fuels throughout the world. High burning and strong radioactive spent fuel resulting from the evolution of various reactors drive the development of the advanced PUREX technology, which emphasizes the separation of neptunium and technetium besides the separation of the Uranium and Plutonium from the majority of highly active fission products. In addition, through Partitioning and Transmutation method, some benefits such as segregating the actinides and long life fission products from the high level waste can be obtained. The GANEX process exploited by CEA, which roots in COEX process belonged to advanced PUREX process, considers the separation of the actinides and long life fission products. The study on the pyro-chemical processing such as the method of electro-deposition from molten salts has still not replaced the traditional PUREX process due to various reasons. In conclusion, the future PUREX process will focus on the modified process including predigesting the technical flowsheets and reducing reprocessing costs and using salt-less reagent in order to minimize the waste production.

Commentary by Dr. Valentin Fuster
2010;():469-475. doi:10.1115/ICONE18-29751.

Several experiments have been carried out with a Nd-YAG laser as a future dismantling cutting tool. At the beginning, industrial lasers could cut only small thicknesses due their small power (1 kW) but now, with the rise in industrial power (8 kW), it is possible to cut plates up to 100 mm thickness. This technique is practical with the use of optical fibers which allow to maintain the laser generator in a non-radioactive zone. This paper provides a synthesis of the measurements of aerosols and gases produced in different configurations (cutting in air and underwater) and some comparisons with other cutting tools.

Commentary by Dr. Valentin Fuster
2010;():477-482. doi:10.1115/ICONE18-29875.

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfill the needs for new transport or storage casks designed to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. In this presentation we will focus on the casks used to transport the fresh and used MOX fuel. To transport the fresh MOX BWR and PWR fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimize the loading in terms of integrated dose and also of course capacity. MOX used fuel has now its dedicated cask: the TN112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the EDF nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.

Topics: Fuels
Commentary by Dr. Valentin Fuster
2010;():483-488. doi:10.1115/ICONE18-29889.

The poolside examination is demanded essentially for the development of an advanced fuel, because it provides data about in-reactor fuel behaviors as well as product performance. Therefore, all the results of examinations are transmitted back to design groups, and evaluated with the design models and codes applied new fuel development. In general, PSE is performed a variety of examinations in assembly and single rod state, such as visual inspection, fuel assembly length growth, bowing and twist, rod-to-rod spacing, spacer grid width, fuel rod diameter, and fuel rod oxide thickness at the end of each cycle of irradiation during plant outage periods. After the completion of fuel life, selected rods are extracted from the fuel assembly for individual fuel rod measurement. In this paper, the techniques of PSE applied an advanced pressurized water reactor fuel, PLUS7™ for Optimized Power Reactors (OPR 1000s) in Korea are described.

Commentary by Dr. Valentin Fuster
2010;():489-499. doi:10.1115/ICONE18-29914.

Reprocessing of used nuclear fuel and treatment of nuclear waste are important issues for the sustainable development of nuclear energy. It is necessary to develop novel nuclear waste treatment technologies to meet the goal of minimizing the secondary liquid waste. Supercritical fluids are considered green solvents in chemical engineering process. It gains growing interest to treat nuclear waste using supercritical fluid extraction recently, because it can greatly decrease the secondary liquid waste with high radioactivity. During the past two decades, extraction of actinides and lanthanides by supercritical fluid has been intensively studied in some countries, and many important progresses have been made. However, the prospect of industrial application of supercritical fluid extraction technology in reprocessing of used nuclear fuel and treatment of nuclear waste is still unclear. In this paper, extraction of actinides and lanthanides from various matrixes or from their oxides by supercritical fluid including the experimental results, extraction mechanism and kinetic process was reviewed. The engineering demonstration projects were introduced. The trend of industrial application of supercritical fluid extraction technology in nuclear waste management was also discussed.

Commentary by Dr. Valentin Fuster
2010;():501-509. doi:10.1115/ICONE18-29985.

A large number of nuclear reactors with graphite as moderator and reflector material are facing to be decommissioned now or later, and the radioactive graphite waste is a large part of the involved wastes. In addition, high temperature gas-cooled reactors being developed rapidly use a large quantity of graphite material (up to 95%) in the nuclear fuel elements, besides graphite material as their moderator and reflector material in the reactor cores. Therefore, it is very critical to manage these graphite wastes from the decommissioned and being decommissioned reactors. The part with low-level radioactive contamination that could not be reused now, may be disposed of as solid waste to reduce its volume, and the possibility of its being retrieved and reused in the future with advanced technology should be considered. The other graphite waste with high-level radioactive contamination requires much more consideration. Due to several factors, such as its large quantity, a lack of available disposal sites and public acceptance, it may not be disposed of directly in the repository any more. An option may be the transformation of the high-level radioactive graphite waste into low-level radioactive waste through physical and chemical processes. The current technologies involve, e.g., thermal treatment to release 36 Cl, capture of the 14 C from the gases of incineration of carbon material and decomposition of carbon dioxide into solid carbon. After these treatments the carbon material might be decontaminated and separated as low-level radioactive waste and a small amount of residual high-level waste could be disposed of ultimately. In order to achieve a sustainable development of graphite material, the maximum utility and the minimal disposal of radioactive graphite should be considered in the management of radioactive graphite waste. It is urgent to explore new technologies for decontaminating and recycling radioactive graphite.

Topics: Graphite
Commentary by Dr. Valentin Fuster
2010;():511-515. doi:10.1115/ICONE18-30008.

Reduction of radioactive waste volume is one of the important issues for spent fuel reprocessing plants. JGC has intensively developed radioactive waste processing technologies aiming at waste volume reduction and at immobilization with environmentally low impact for Purex type reprocessing plants. These technologies are Electrolytic Decomposition Process of Nitric Acid (E-DeP), Catalytic Decomposition Process of Nitrate Ion (C-DeP) and Electrodialytic Na Separation Process (E-SeP). An effective combination of each technology results in reduction of numbers of waste packages by one-twentieth.

Commentary by Dr. Valentin Fuster
2010;():517-521. doi:10.1115/ICONE18-30041.

Core loading pattern design has great influence on nuclear power plant operation. An excellent core loading pattern can not only enhance operation factor, reduce operation cost, but also increase operation safety. Under the premise of nuclear safety, AP1000 first core loading pattern achieves the goal of low leakage loading by simulating the reactivity distribution of the 18-month Equilibrium Cycle design. The fuel management presented in this paper illustrates the economic performance and technical feasibility of the advanced 18-month cycle first core fuel. The advanced feature of this first core design include: the development of loading pattern that simulates the reactivity distribution of a typical low leakage reload core in order to reduced leakage, the use of radial enrichment zoning in higher enriched assemblies to lower peaking factors, and the use of axial burnable absorber zoning to improve axial power shape control and reduce axial peaking. The discussion provided in this paper demonstrates the ability of the advanced first core design operate safety and efficiently, and the core is designed with adequate peaking factor margin in both base-load and load-follow operation. Finally, this paper analyses the impact brought about by multi-enrichment on the nuclear power plant operation capacity and operation cost, and arrive at a conclusion that precise fuel cycle evaluation such that the enrich uranium and operation costs can be accurately quantified and control in a detailed economic evaluation.

Commentary by Dr. Valentin Fuster
2010;():523-527. doi:10.1115/ICONE18-30092.

The removal efficiencies of strontium, cesium and cobalt by inorganic membrane with cut-off 8kD were studied. In order to improve the removal efficiency of non-active nuclides, soluble macropolymers sodium poly(acrylic) acid (NaPAA) of different molecule sizes were selected as chemical assistant reagents to make comparative experimental studies. The molecule of 8000,50000 and 100000 D NaPAA were used. The flux and removal rates of different systems were mainly studied. The results show that for inorganic membrane with pore size of 8kD molecule of 50000 Da NaPAA is the best selection for assistant reagents.

Commentary by Dr. Valentin Fuster
2010;():529-532. doi:10.1115/ICONE18-30110.

This paper presents the design of the low-level solid waste encapsulation device for the 200L barrels. The device is used to compress and package the low-level contaminated waste such as cotton goods, rubber, plastic products, leather products and paper products.

Topics: Solid wastes , Design
Commentary by Dr. Valentin Fuster
2010;():533-543. doi:10.1115/ICONE18-30123.

With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU® reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past [1–4]. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100 to 1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

Commentary by Dr. Valentin Fuster
2010;():545-549. doi:10.1115/ICONE18-30137.

The R&D of coating process of fuel particles is one of the most important key technologies in the research work of high-temperature gas-cooled reactor (HTGR). A safe and reliable coating process is expected and related to the prospect of large-scale utilization of nuclear energy. The related research of the carbon black powder which is the main byproduct generated in the coating process is important, because it relates to the impact of coating process on the surrounding environment and is also helpful to understand the deposit mechanism of PyC (pyrolytic carbon) layer coated on the fuel particle. In the present study, the microstructure of the pyrolytic carbon powder were systematically investigated by scanning electron microscope (SEM), transmission electron microscopy (TEM), laser Raman spectroscopy and particle size analysis (PSA). It can be found that the carbon powder in the cyclone separator is composed of the nano-spherical particles with the diameter of about 50nm. The ring-layered nano-structured carbon particles could be found from the electron micrographs. The comparison between Raman spectra of carbon powder and pyrolytic carbon coated on the fuel particle showed that the droplet deposit mechanism was suitable to explain the PyC deposition process. The nano-particles agglomerate into the irregular floc and the diameters of the stable particle clusters are mainly 1 μm and 10 μm. The disposal methods of carbon black powder are also discussed.

Commentary by Dr. Valentin Fuster
2010;():551-557. doi:10.1115/ICONE18-30141.

As an alternative grouting material for the geological repository of long-lived radioactive waste, the “S uperf ine S pherical silica G rout” (SFSG) material is developed using a fine spherical silica and a fine calcium hydroxide. The developed SFSG material takes an advantage of its smaller particle size distribution (max. ∼1 micron or less) than those of the cementitious materials, and also provides a low alkaline environment so as to reduce unfavorable effects on the long-term performance of geological disposal system. The SFSG is a mixture of the “super fine silica powder”, the superfine calcium hydroxide and additives such as superplasticizer. Presently, the mixture being investigated for grouting materials is focused on water/binder ratio (W/B) of 1.2. Some preliminary laboratory experiments were carried out to characterize its fundamental properties from the viewpoint of practical use for geological disposal, which is required to be equivalent with the conventional cementitious materials in terms of penetrability, strength, pH performance and workability. From a series of experiments, it was concluded that SFSG is expected to become an alternative grouting material for a geological repository.

Commentary by Dr. Valentin Fuster
2010;():559-564. doi:10.1115/ICONE18-30154.

To investigate the solidification efficiency of sulfoaluminate cement (SAC) and to provide more information for formula optimization, SAC blending zeolite, accelerator and Dura fiber was used as matrix materials for solidification of simulated radioactive borate liquid waste. The simulated radioactive borate liquid waste was prepared with boric acid and sodium hydroxide using drinking water. The performances of solidified waste forms were evaluated mainly basing on matrix compressive strength and leachability. The 28d compressive strength of the solidified waste forms were tested according to Chinese National Standard GB 14569.1-1993, and experiments on water/freezing/irradiation/impact resistance were also carried out. Nuclides Sr, Cs and Co were substituted by their non-radioactive isotopes respectively in leachability test, and the testing procedures were consistent with Chinese National Standard GB 7023-1986. Experimental results showed that it was feasible to solidify borated liquid wastes with SAC. The 28d compressive strength was 13.9MPa, nearly twice of the standard in GB 14569.1-1993. Strength losses in water/freezing/irradiation/impact resistance tests met the demands of GB 14569.1-1993 well. In the leaching test, the 42d leaching rates were 3.39×10−5 cm/d, 4.45×10−5 cm/d and 4.07×10−7 cm/d for Sr2+ , Cs+ and Co2+ respectively, much lower than GB 14569.1-1993 limits. Results of leaching test also showed that the leaching mechanism of Co2+ was different from that of Sr2+ and Cs+ .

Commentary by Dr. Valentin Fuster
2010;():565-570. doi:10.1115/ICONE18-30164.

As a lower cost raw material, few demand for equipments, convenient solidification process, cement solidification for radioactive waste is widely used for several decades. Formulations of solidification are complex and diverse, involving various types of substrate and additives. Traditional approach for formulation design is single-factor test whose representation is inadequate and workload is huge. Uniform design based on the theory of Quasi-Monte-Carlo takes advantage of limited and representative tests instead of the system. In the multi-factor formulation design, it can be very quick and convenient to find the formulations required by uniform design table and direct-vision method. This article introduced the application of uniform design table for formulation in cement solidification of nuclear waste resin.

Commentary by Dr. Valentin Fuster
2010;():571-575. doi:10.1115/ICONE18-30166.

In Japan, the clearance system has been in effect since December 2005, and at present, the cleared material is recycled at a site in affiliation with the nuclear installation. In the near future, it is assumed that a considerable amount of cleared material will be transported from nuclear facilities that are being decommissioned, so that the material can be recycled and used by general industry. In order to operate the clearance system with certainty, it is essential to devise a method for promptly measuring the radionuclide concentration of cleared materials that are suspected of exceeding the clearance level. That is, preparation for the contingency that an uncertain radiation source mixes with a clearance item by chance. In general, in-situ measurement of low-level radionuclide concentration is difficult since accurate information on the shape of the target, as well as on the distribution of radionuclide concentration in the target, is required. Therefore, we investigated a method for simple and rapid determination of radionuclide concentration, which does not require accurate information on the volume and density of the target. To avoid the influence of volume or density, we examined the adoption of a conversion factor with units of “(Bq/g)/cps”. ISOCS (In-situ Object Counting System) Calibration Software (Model S573 Version 4.0, CANBERRA Inc.) was used to determine the conversion factor. Metal or concrete was assumed to be stored and piled in a stockyard of a recycling facility. The piled objects were arranged as a cone with a gradient of 50%, in accordance with the limitations specified by the “Waste Management and Public Cleansing Law”. The volume and bulk density were assumed to be 0.1–10 m3 and 0.5–10 g/cm3 , respectively. The conversion factor for a cone with volume of 1 m3 and density of 1 g/cm3 could be applied to the piled objects having a volume greater than or equal to 1 m3 and a density greater than or equal to 0.5 g/cm3 , resulting in an increase in tolerance by a factor of two. These results show that in-situ measurements of radionuclide concentration can be simplified by using a conversion factor with units of “(Bq/g)/cps”. This approach can be used to carry out prompt measurements on suspected materials in the field.

Commentary by Dr. Valentin Fuster
2010;():577-580. doi:10.1115/ICONE18-30227.

The GE Hitachi Nuclear Energy (GEH) nuclear criticality safety (NCS) function remains actively engaged in advancements to the nuclear fuel cycle. In addition to its traditional BWR fuel manufacturing, recent GEH emphasis to become more vertically integrated into front end (enrichment) and back end (reprocessing) fuel cycle technologies has had a dramatic impact on the NCS function. Required fundamental and practical research in various fields, such as general physics, computational methods, validation methodology, cross-section data processing, criticality safety assessments, risk-informed integrated safety analyses, and domestic and international nuclear packaging licensing, collectively present significant challenges to NCS staff. As the landscape of the GEH business growth opportunities continues to evolve over time, so does the required depth of NCS knowledge and technical expertise. This paper provides an overview of select NCS design, licensing, methods, and packaging activities in support of GEH nuclear fuel cycle business subsidiaries and concludes with some insight to technical and regulatory challenges.

Commentary by Dr. Valentin Fuster
2010;():581-583. doi:10.1115/ICONE18-30336.

The ROXf is a kind of inert matrix fuel U-238-free matrices, it has a high plutonium transmutation capability. The ROXf consists of chemically stable phases of fluorite ‘stabilized’ ZrO2 or ThO2 , and spinel MgAl2 O4 . In this fuel, PuO2 is solidified in a fluorite phase. With U-238-Free matrices, a large part of the plutonium can be burned after irradiation in conventional LWRs. The spent ROXf consists of natural analogous geologically stable phases, and is disposed directly as high level wastes ‘HLWs’ after about 50 years cooling. From the high plutonium burn up rate and the high stability of the fuel, the ROXf-LWRs system has proliferation resistance and environmental safety. Characteristics of two types of ROXf, Zr-ROX and Th –ROX with weapons-Pu, in an LWR core arrangement are evaluated by cell burn up calculations and 2-D core calculations using the SRAC code system and JENDL-3 nuclear library. In an LWR of moderator to fuel volume ratio = 1.9, which corresponds to current PWRs, Pu transmutation rates with the two types of ROXf are large enough and more than 80% and 99% of Pu and Pu-239, respectively, can be burned. The calculated kinetic parameters indicate less moderate characteristics of ROX cores, especially with Zr-ROX. The fertile Th-232 in Th-ROX works like U-238 in the UO2 fuel, making kinetic parameters more moderate and reactivity drop due to burn up smaller than that Zr-ROX. The neutron capture of Th-232 to generate U-233, causes the safeguards problem. Thus, the characteristics of Zr-ROX as a typical example, were investigated in this study.

Topics: Combustion , Fuels , Rocks
Commentary by Dr. Valentin Fuster

Instrumentation and Controls (I&C)

2010;():585-592. doi:10.1115/ICONE18-29001.

The Core Protection Calculator System (CPCS) was the first implementation of digital computers in a nuclear power plant safety protection system. The system was based on first principles to calculate the specified acceptable fuel design limit (SAFDL) online. This approach provides the theoretical optimum safety margin. The first-of-its-kind system was installed in the United States at Arkansas Nuclear One Unit 2 (ANO-2) in 1980. Extensive efforts were made by Combustion Engineering and U.S. Nuclear Regulatory Commission (NRC) staff to gain licensing approval of the CPCS. Based on accumulated operating experience, numerous improvements were made to enhance the performance of the CPCS. The CPCS software provided the flexibility to readily accommodate these design changes. Currently, CPCS is implemented in 21 nuclear power plants in operation or under construction in the U.S.A. and Asia. The next generation CPCS will focus on optimizing the plant protection by improving the SAFDL calculation. By taking advantage of the advances in digital computer technology, the comprehensive safety analysis code will be used online. A more detailed core power map using the incore detector signals will be used as the basis of the departure from nucleate boiling ratio (DNBR) and local power density (LPD) calculation. A quick power reduction will provide adequate margin for most of the design basis events. For these events, CPCS will initiate a reactor power cutback as opposed to a reactor trip, which will maintain the plant at a safe condition with a reduced power level.

Commentary by Dr. Valentin Fuster
2010;():593-598. doi:10.1115/ICONE18-29011.

This paper focuses on the technical improvements for Human System Interface (HSI) implemented to be designed to manage normal and accidental situation of the Nuclear Power Plants (NPPs) on the LING AO 3&4 nuclear plants project under construction in the South of China. Regarding the operation principles of the NPPs, two major improvements on the LAO 3&4 NPPs are introduced: Implementation of a Digital Control System (DCS) combined with a computerized Human System Interface and backed-up with a conventional control mean Back-up panel (BUP). Some technical improvements for HSIs such as State Oriented Procedures (SOP), Large Display Panel (LDP), Computerized-base procedures, Advanced alarm system, Safety Parameter Display system (SPDS) are detailed in this paper. Finally, in the scope of these studies, the human factors considerations are considered in order to reduce the likelihood of human errors, to gain maximum benefit of the implemented technology and to increase the performance.

Commentary by Dr. Valentin Fuster
2010;():599-609. doi:10.1115/ICONE18-29029.

The digitalized Instrumentation and Control (I&C) system of nuclear power plants (NPP) could provide operator easily Human-Machine Interface (HMI) and more powerful overall operation capability. However, some software errors may cause a kind of Common Cause Failure (CCF). As a consequence, the event of Anticipated Transients Without Scram (ATWS) will occur. In order to assure that the plant can be shutdown safely and to follow the requirements of 10CFR50.62, the utility builds up various ATWS mitigation features in NPP. The features include Fine Motion Control Rod Drive Run In, Alternate Rod Insertion, Standby Liquid Control System, Reactor Internal Pump Trip or Runback, Feedwater Flow Runback and Inhibition of Automatic Depressurization System. This research developed an evaluation method of diverse back-up means for computerized I&C system. A diverse backup of digital I&C system is the most important means to defend against CCF and un-detectable software faults. Institute of Nuclear Energy Research (INER) is developing a computerized I&C test facility, which is incorporated a commercial grade I&C systems with Personal Computer Transient Analyzer (PCTran)/Advanced Boiling Water Reactor (ABWR), a NPP simulation computer code. By taking the technology of Field Programmable Gate Array (FPGA) to implement the methods of ATWS mitigation, the research built up a diverse back-up of digital I&C system to expect to defend against CCF and undetectable software faults. According to the testing and evaluation, the work can be achieved the analysis of Diversity and Defense-in-Depth (D3).

Commentary by Dr. Valentin Fuster
2010;():611-618. doi:10.1115/ICONE18-29055.

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

Commentary by Dr. Valentin Fuster
2010;():619-626. doi:10.1115/ICONE18-29056.

Due to the existing serious climate and environment problems caused by burning fossil fuels, nuclear energy is now rapidly developed. Power-level control for nuclear reactors is significant for not only regular operating but also safety issues. The advances in computer technology, information processing, and control theory in the past decades allow the applications of advanced controllers with higher performance. In this paper, the separation principle for the recently established dissipation based high gain filter (DHGF) is presented, which guarantees the closed-loop stability of the system interconnected with a state-feedback controller and the DHGF. Then, the feasibility of applying the DHGF to the load-following control of nuclear reactors has then been verified. Finally, the DHGF with a well designed state-feedback power-level controller has been successfully applied to realize the load-following control for a nuclear heating reactor (NHR). Numerical simulation results has shown the high performance of the DHGF and the feasibility of the DHGF-based output feedback control strategy, and influence of the observer gain to the control and observation performance has also been analyzed.

Commentary by Dr. Valentin Fuster
2010;():627-633. doi:10.1115/ICONE18-29089.

Liquid level measurements have been studied for a long time. Many research systems have been constructed. Some sophisticated systems have successfully experiences in industrial applications for years. But, in some specific situations, the liquid level is replaced by multiphase media level, such as bubbly flow level or foam level. Those have been widely observed in waste disposal process, chemical engineering fields and even during process of Lost Of Coolant Accident (LOCA) events in the Pressure Water Reactor (PWR). As the foam is a special phase compared with liquid, gas or steam, the unique mechanics, thermodynamics and optics characteristics make the measurement for it much difficult. It is still a Cutting-edge topic in level measurement field by now. A unique foam level sensor is described in this paper. The structure, principle are analysis in detail.

Topics: Sensors
Commentary by Dr. Valentin Fuster
2010;():635-640. doi:10.1115/ICONE18-29102.

As Nuclear Power Plant (NPP) adopted software-based Digital Instrumentation and Control (DI&C) System, the Software Configuration Management (SCM) is becoming more and more important for NPP. The main reason is the inherent changeability and invisibility which of software often causes unpredictable problems and results are difficult to manage. In addition, the DI&C system has always been constructed by multiple vendors and each vendor of the DI&C has its own development artifacts. Therefore, it is great challenge for NPP staff to maintain the consistency and integrity of software Configuration Items (CI) among multiple vendors. The software CIs include software products delivered to customers and items required to create the software products such as software design document, source code, database, test report, compiler, etc. In general, Software Configuration Management System (SCMS) is usually developed to support SCM activities, such as storing CI, controlling change, and accounting and auditing throughout the entire software lifecycle. However, most existing file-based SCMS typically deal with those artifacts of individual files without providing the more detailed configuration and change information among CIs. Based on the nuclear SCM related regulations, this paper proposes a developing SCMS for the DI&C system of a NPP. Its main goal is to meet the regulatory requirements, and enhance the visibility, tractability and integrate ability to manage the heterogeneous subsystems within the DI&C system. This paper provides the more detailed descriptions about regulation requirements analysis, system design and the development process. Finally, a prototype system is presented.

Commentary by Dr. Valentin Fuster
2010;():641-643. doi:10.1115/ICONE18-29224.

For the special nature of the nuclear physics experiment, a NIM-standard nuclear instrumentation system is simulated based on LabVIEW, including plug-ins simulation and integrated devices such as a nuclear random signal generator, a linear amplifier, a single channel analyzer, a scaler and a data-processing module. In order to verify the system, some undergraduate nuclear physics experiments were done. The results show that the simulation signals of the nuclear random signal generator accord with the rule of statistical nature of the nuclear random pulse, the GUI of simulation system is verisimilar, simulation operation is basically the same as real experimental operation. Application of the simulation system has practical significance in mastering conventional nuclear radiation detection methods and avoiding radiation damage of the experiments.

Commentary by Dr. Valentin Fuster
2010;():645-649. doi:10.1115/ICONE18-29230.

An open-walled ionization chamber is developed to monitor the tritium concentration in glovebox in the tritium processing systems. Two open walls are used to replace the sealed wall in common ionization chambers, through which the tritium gas can diffuse into the chamber without the aid of pumps and pipelines. Some basic properties of the chamber are examined to evaluate its performance. Results turn out that an open-walled chamber of 1 L in volume shows considerably flat plateaus over 700 V. The chamber also gives a good linear response to the gamma field over five degrees under condition of 1 atm. The pressure dependence characters show that the ionization current is only sensitive to low pressure. The pressure influence becomes weaker as pressure increases mainly due to the decrease of the mean free path of β ray decayed by tritium. The minimum detection limit of the chamber is 3.7×105 Bq/m3 .

Commentary by Dr. Valentin Fuster
2010;():651-657. doi:10.1115/ICONE18-29232.

In this paper, we develop a monitoring system of reactor coolant pumps in nuclear power plant (CPS). The safe running of reactor coolant pump is important for nuclear power plant. Based on the Fourier transform (FT) and some algorithm, The data collected from the pump are analyzed. Once the accident happens, it would cause unimaginable outcome. The system will be jumped to failure process mode when the pump has something wrong. The advanced VXI and virtual instrument technology are applied to system, and the reactor coolant pump will be monitored overall so as to assure that the reactor coolant pump runs in safe, which has a significant value to secure the safe operation and reliability of the nuclear plant. The monitoring system will help the operators find fault of reactor coolant pump.

Commentary by Dr. Valentin Fuster
2010;():659-665. doi:10.1115/ICONE18-29264.

All 1000 MW nuclear power plants currently in construction or projected to-be-built in China will use the digital instrumentation and control (I&C) systems. Safety and reliability are the ultimate concern for the digital I&C systems. To obtain high confidence in the safety of digital I&C systems, rigorous software verification and validation (V&V) life-cycle methodologies are necessary. The V&V life-cycle process ensures that the requirements of the system and software are correct, complete, and traceable; that the requirements at the end of each life-cycle phase fulfill the requirements imposed by the previous phase; and the final product meets the user-specified requirements. The V&V process is best illustrated via the so-called V-model. This paper describes the V-model in detail by some examples. Through the examples demonstration, it is shown that the process detailed in the V-model is consistent with the IEEE Std 1012-1998, which is endorsed by the US Regulatory Guide 1.168-2004. The examples show that the V-model process detailed in this paper provides an effective V&V approach for digital I&C systems used in nuclear power plants. Additionally, in order to obtain a qualitative mathematical description of the V-model, we study its topological structure in graph theory. This study confirms the rationality of the V-model. Finally, the V&V approach affording protection against common-cause failure from design deficiencies, and manufacturing errors is explored. We conclude that rigorous V&V activities using the V-model are creditable in reducing the risk of common-cause failures.

Commentary by Dr. Valentin Fuster
2010;():667-671. doi:10.1115/ICONE18-29286.

Reactor protection system is one of the most important safety systems in nuclear power plant and shall be designed with very high reliability. Digital computer-based Reactor Protection System (RPS) takes great advantages over its conventional counterpart based on analog technique and faces the issues how to effectively demonstrate and confirm the completeness and correctness of the software that performs reactor safety functions in the same time. It is commonly accepted that the essential way to solve safety software issues in a digital RPS is to pass a strict and independent Verification and Validation (V&V) process, in which integrated RPS testing play an important role to form a part of the overall system validation. Integrated RPS testing must be carried out rigorously before the system is delivered to nuclear power plant. The integrated testing are often combined with the factory acceptance test (FAT) to form a single testing activity, during which the RPS is excited by emulated static and dynamic input signals. The integration testing should simulate normal operation, anticipated operational occurrences and accident conditions, as well as anticipated faults on the inputs to the DRPS such as sensors out of range or ambiguous input readings. All safety function requirements of digital RPS should be confirmed by representative testing. The design and development of a test facility to carry out the integrated RPS testing are covered in this paper, which is merged in the research on a digital RPS engineering prototype for a nuclear power plant. The test facility is based on PXI platform and LabVIEW software development environment and its architecture design also takes into account the test functions future extensions such as hardware upgrades and software modules enhancement. The test facility provides the digital RPS with redundant, synchronized and multi-channel emulated signals that are produced to emulate all protection signals from 1E class sensors and transmitters with time varied value within their possible ranges, which would put integrated RPS testing into practice to confirm the digital RPS has fully met its predefined safety functionality requirements. The designed test facility can provide an independent verification and validation process for the research of digital RPS with scientific methods and authentic data to evaluate the RPS performance thoroughly and effectively, such as measuring threshold precision and trip response time, analyzing system statistical reliability and so on.

Commentary by Dr. Valentin Fuster
2010;():673-677. doi:10.1115/ICONE18-29305.

Obsolescence presents great challenge to Nuclear Power Plants (NPP) and plant simulators around the world. Old designs will have to be either modified, or replaced by new designs, in order to simplify maintenance, increase availability and meet ever-increasing operational and training requirements. Control system upgrade and Distributed Control System (DCS) design for both old plants and new builds have become the center of interest. In a DCS, communication networks connect control systems together to allow the exchange of information and feedback. Among the many communication network protocols, Ethernet can be a promising one. This paper describes a new Ethernet Bus Interface Controller (eBIC) used in the Input/Output (I/O) system of a Nuclear Power Plant (NPP) simulator in Canada.

Commentary by Dr. Valentin Fuster
2010;():679-684. doi:10.1115/ICONE18-29310.

This paper investigates the minimum inventory (MI) of human system interfaces (HSIs) (i.e. alarms, controls, and displays) for plant’s safe operation and represents the analytic procedure on the MI of HSIs developed for the digital instrumentation and control (I&C) equipments in the main control room (MCR). The MI of HSIs in the MCR indicates the HSIs that the operator always needs available to: (1) monitor the status; (2) perform and confirm a reactor trip; (3) perform and confirm a controlled shutdown of the reactor; (4) actuate safety related systems; (5) analyze failure conditions of the normal HSIs; (6) implement the plant’s emergency operating procedures (EOPs); (7) bring the plant to a safe condition; (8) carry out those operator actions shown to be risk important by the probabilistic risk assessment (PRA). The proposed analytic procedure on the MI of HSIs in this study can be used to (1) identify the MI of HSIs and their design requirements; and (2) addresses design requirements and implementation for the MI of HSIs. The contribution of this study is to describe the MI of HSIs needed to implement the plant’s EOPs, to bring the plant to a safe condition, and to carry out those operator actions shown to be risk important by the PRA.

Commentary by Dr. Valentin Fuster
2010;():685-690. doi:10.1115/ICONE18-29396.

The nuclear industry and research institutes in Taiwan are conducting a joint effort project to establish a self-reliant nuclear Instrumentation and Control (I&C) system design and fabrication capabilities in Taiwan. The purposes of this project, as called Taiwan’s Nuclear I&C System (TaiNICS), are planned to support digital upgrade of the existing nuclear power plants and the new nuclear installations in Taiwan. The project will be a long term pursuit of several task branches, including establishment of a generic qualified digital platform, qualification and certification processes, nuclear I&C systems design, safety analyses for software common cause failure, licensing, and collaboration. The short term goal of this project is to submit the License Topical Report (LTR) of a generic digital platform for the review of Taiwan’s regulatory body in 2013.

Commentary by Dr. Valentin Fuster
2010;():691-699. doi:10.1115/ICONE18-29487.

The paper introduces the first full digital Instrumentation & Control System (referred to as “I&C System” hereinafter) in Tianwan Nuclear Power Station in China, in which the computer and Ethernet network were adopted as the technique platform. After introduction of the characteristics and composing of the I&C system, some problems appeared during commissioning and normal operation are described, mainly focuses on I&C system design optimization, major defects elimination and solution, reasons analysis and corrective remedies for typical events. Since the two units were in commercial operation in 2007, “nuclear safety culture” of the station has been promoted persistently, management and administration regulations are strict and precise, responsibilities are clearly assigned, and documentation system is completely developed, work is being continuously improved. The full digital I&C system is maintained by qualified I&C staffs strictly in accordance with the procedures; The full digital I&C system has exhibited its excellent qualities and remarkable characteristics as follows: stable operation, reliable function, friendly man-machine interface, intuitionistic operation and monitoring, easy configuration, and convenient trouble-shooting. Tianwan Nuclear Power Station I&C system has never caused any event that led to change of the unit operation status, thus has contributed great deal to the safe, stable and economic operation of the two units.

Commentary by Dr. Valentin Fuster
2010;():701-706. doi:10.1115/ICONE18-29500.

A remote control system is designed to achieve the stable and economical operation of far-infrared laser interferometer on the EAST Machine (Experimental Advanced Superconducting Tokamak). Based on the PLC controller, the system is set up as two parts: gas flow control for adjusting working gas, feedback control for stabilizing output power of the laser. We demonstrate the feasibility that, according to the need of experiment, the working gas ratio mode can be changed in order to save gas (CD4 ) instead of keeping a single-mode ratio. Meanwhile, the changes of the output power and power supply of the laser have been measured due to the mode conversion. The output power of the DCN laser, as a laser source, is influenced by a number of parameters, especially the gas ratio and a variety of disturbances. Therefore, the feedback control system has been developed to make the laser power stable near the maximum available power. The whole system has been applied to the DCN laser, and has been proved to be effective.

Commentary by Dr. Valentin Fuster
2010;():707-710. doi:10.1115/ICONE18-29510.

The technology of real-time fault diagnosis for NPP has great significance to improve the safety and economy of reactor. At present, expert system, artificial neural network (ANN) and support vector machine (SVM) algorithms are most widely used in the field of NPP fault diagnosis. According to the shortcomings of expert systems, ANN and SVM, the decision tree algorithm is applied in the field of NPP fault diagnosis in this paper. ID3 and C4.5 are applied separately to learn from training samples which are the typical faults of NPP, and diagnose using the acquired knowledge. Then the diagnostic results are compared with the results of SVM method. The results show that: comparing with SVM, decision tree has the advantages of much faster training speed and a little higher accuracy. Furthermore, decision tree can obtain rules from the sample set, so it has good explanatory ability for the diagnostic results.

Commentary by Dr. Valentin Fuster
2010;():711-713. doi:10.1115/ICONE18-29512.

Usually the plus of the voltage-preamplifier we used is changeless or 2-modes (×1 or ×10), and the value is low (×10). But in the practical application, the energy of the particles detected are complicated. Especially, when the particles which are detected are unknown, it will be advantageous to the result to select a proper plus. So it will be great beneficial for the practical application to design a bran-new voltage-preamplifier which has bigger plus comparing to the amplifier we used now and could select a proper plus for the experiment automatically. This article introduces a design of a kind of new voltage-preamplifier with a bigger plus (×20) than traditional voltage-preamplifiers and whose plus can be controlled continuously. The most important of all is that this voltage-preamplifier can select a appropriate plus for the experiment automatically by the order of the handlers. It adjusts the plus by the information from the sensor to the order. In this design we control the output-signal from 100∼200 mV. And then, this paper introduces the simulation of the amplifier’s working process by Multisim10.0 and Multisim10.0 Proteus ISIS.

Commentary by Dr. Valentin Fuster
2010;():715-721. doi:10.1115/ICONE18-29556.

Main control room in nuclear power plants (NPP) is a complex system where operators interact with a large amount of human system interface (HSI) resources, which is essential to the safety of the plant. In order to achieve a high standard of human factors engineering (HFE) level and ensure the effectiveness and efficiency of the system, verification and validation (V&V) should be performed before the delivery of the plant. This study firstly represents an overview of the HFE project and V&V activities applied in the nuclear industry and relative researches, then, a V&V program from an ongoing 300MW NPP project is discussed in detail. Comparative methods, existing system vs. design requirements, are mainly applied for the verification phase. In the HFE Design Verification, 10 different HFE design guideline sets and corresponding checklists are developed for the experienced reviewers to conduct a thoroughness and consistency review. Critical safety functions, risk important tasks, critical human actions and other necessary operations are verified in the Task Support Verification to insure that HSI provides all required resources for personnel tasks. Other than direct comparison, a development platform is also used to assist the analysis of display, alarm and other appropriate features. In the validation phase, an integrated system test would be completed on a full scope simulator with the participation of experienced operators from the utility as subjects and multi-discipline observers. Scenarios are carefully designed to ensure the representative of the test. Both objective and subjective results would be collected and processed mainly by the descriptive statistics method to find out the problems in the existing design. All the discrepancies found in the whole V&V would be involved in a specific database and tracked until resolved. Lessons learned from this case are discussed.

Commentary by Dr. Valentin Fuster
2010;():723-730. doi:10.1115/ICONE18-29587.

The reviewing of operating experience at nuclear power plants (NPP) is not only critically important to safe and reliable operations, but also useful to guide the design of new plants which are similar to the current one under review. How to identify and analyze the safety-related operating experience and then implement a more extensive review is a vital and challengeable issue. In this paper, a methodology of human factor engineering (HFE) operating experience review (OER) is proposed for NPP. The need for the application of HFE in the life cycle activities of NPP and other nuclear facilities has been demonstrated by plant operating histories and regulatory and industry reviews. As a very important element of HFE, the OER is performed from the beginning of the design process. The main purpose of performing an OER is to verify that the applicant has identified and analyzed HFE-related safety problems and issues in previous designs that are similar to the current one. In this way, negative features associated with predecessor designs may be avoided in the current NPP design while retaining positive features. The research of OER concentrates on the aspect of review criterion, scope and implementation procedure of the HFE-related operating experience. As the NRC requirement, the scope of operating experience can be divided into six types in accordance with sources of information. The implementation procedures of USA and China are introduced, respectively. The resolution of HFE OER issues involve function allocation, changes in automation, HSI equipment design, procedures, training, and so forth. The OER conclusions can contribute to other HFE activities and improve the safety, reliability and usability of the HSI design in NPP.

Commentary by Dr. Valentin Fuster
2010;():731-736. doi:10.1115/ICONE18-29589.

Radiation monitoring plays a vital role in the safe and efficient operation of the nuclear power plant (NPP). The current radiation monitoring system (RMS) generally uses cable monitoring network with distributed radiation monitors. It will introduce various compatibility issues that more detector nodes are added to the existing cable monitoring network. The communication protocols from different device manufacturers are not compatible. Furthermore, the original RMS has to be shut down for rewiring and reconstruction. In this paper, a heterogeneous framework is proposed based on the wireless sensor network (WSN) technology, for monitoring environmental conditions around and inside NPP, specifically, radiation levels. The proposed full-scope RMS has a no-wiring and no-construction upgraded scheme based on the WSNs, which forms a heterogeneous multi-networks fusion control system, and does not affect the existing NPP radiation monitoring facilities. The introduction of the wireless gateway to build heterogeneous monitoring framework makes it possible to complete the system seamless upgrade with a lower cost and higher feasibility.

Commentary by Dr. Valentin Fuster
2010;():737-742. doi:10.1115/ICONE18-29596.

Recently Human Reliability Analysis (HRA) is becoming more important to the safety of nuclear power plant (NPP). As the reliability of the NPP equipments have been increased more higher, HRA should be developed in order to guarantee the better safety of NPP. By the collection of human performance about operators in main control room of NPP, especially in accident situation, it is very important to enhance the human reliability. This paper chooses Loss of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) as the initiating events, and base on this, some loss of other equipment or system have been added. Then the process that the operators deal with the accidents has been recorded in the accident situation in order to get reasonable human performance data. After we analyzed all of these video by some tools such as eye tracking tool, some items have been found which are very important to the human reliability such as training level, operation task, human-machine interface, surrounding, team work, etc. According to the analysis of the human performance result, we have evaluated the computerized human machine interface. As a result, it also could be benefit to improve the design of NPP, specially the design of main control room.

Commentary by Dr. Valentin Fuster
2010;():743-748. doi:10.1115/ICONE18-29600.

In the past, the human machine interface is not computerized and the analysis tools are lack. So that most of previous researches of operator performance focused on the performance of individual operator rather than on the team performance. The relationships between team performance and individual performance are very close, but the team performance is not exactly equal to the sum of individual performance. The complexity of team performance evaluation is much higher than that of individual performance evaluation. Along with the development of new technology, especially the computerized technology, such as eye tracking tool and video recorder, team performance could be observed and measured. In NPP (Nuclear Power Plant), the operator performance of team work is closed to the safety of the whole plant. So the operators needed to be paid high attention to team performance. In order to improve the operator performance (including individual and team performance) base on the computerized human machine interface, this study bends to develop a real-time warning model for assessing team performance through the mental workload and human reliability evaluation. According to the analysis result, it could be as reference to modify the design about human machine interface in order to increase system safety and team performance of NPP.

Commentary by Dr. Valentin Fuster
2010;():749-754. doi:10.1115/ICONE18-29603.

Now the new control technology has been used in industry including the main control room design of nuclear power plant (NPP). Although We have designed the computerized human-machine system interface (HSI) in Ling Ao Phase II NPP, the old evaluation method about HSI can not be used for the new one. Based on this experience together with reference to the other published information such as NUREG-0700 Rev.2, this study have developed an evaluation method based on a model of human cognitive processes in order to provide the viewpoint of the evaluation on the operability of interface. The method describes systematically the human error categories and evaluation items for HSI. And the method has been evaluated using the evaluation of experimental results obtained by a simulator equipped with computers and touch operation control panels.

Commentary by Dr. Valentin Fuster
2010;():755-764. doi:10.1115/ICONE18-29754.

Key challenges caused by implementation of diversity-oriented approach and FPGA technology are discussed in context of NPP I&C systems safety. National and international standards containing the requirements to diversity application in NPP I&C systems are analyzed. A few evolution stages of multi-version NPP I&C systems (Reactor Trip Systems) are described taking into account different types of version redundancy (hardware, software, FPGA diversity). Main attention is attended to the methods of increasing tolerance of NPP I&C systems to physical and design faults using multiversion technologies. A life cycle model and multi-version technologies of FPGA-based I&C systems development are analyzed. Implementation results of safety-critical NPP I&Cs developed by RPC “Radiy” using FPGA technology are described. The FPGA-based platform RADIY™ ensures scalability system functions, dependability and diversity. More than 20 different FPGA-based I&C systems were successfully developed, produced and implemented on the NPPs of Ukraine and Bulgaria during last five years.

Topics: Safety
Commentary by Dr. Valentin Fuster
2010;():765-771. doi:10.1115/ICONE18-29777.

In this paper, kernel principal component analysis (KPCA) is studied for fault detection and identification in the instruments of nuclear power plants. We propose to use mean values of the sensor reconstruction errors of a KPCA model for fault isolation and identification. They provide useful information about the directions and magnitudes of detected faults, which are usually not available from other fault isolation techniques. The performance of the method is demonstrated by applications to real NPP measurements.

Commentary by Dr. Valentin Fuster
2010;():773-782. doi:10.1115/ICONE18-29779.

This paper presents the development of a three-dimensional space-time neutronic kinetic modeling of a CANadian Deuterium Uranium (CANDU) reactor for control system design and research, using a modal method. In this method, the reactor space-time neutron flux is synthesized by a time-weighted series of pre-calculated neutron flux modes. The modes are eigenfunctions of the governing neutron diffusion equation during reference steady-state operation. The Xenon effect has also been considered. The reactor model is then implemented within a simulation platform of CANDU6 reactor regulating system (RRS), in MATLAB/SIMULINK. Non-dimensionalized SIMULINK representation of the reactor kinetic modeling is established. Behavior of the reactor during a load following transient has been simulated using the developed reactor-modeling module. The simulation results prove the efficiency of the reactor modeling. Real-time three-dimensional neutron flux distribution during the transient analysis is represented.

Topics: Spacetime , Stress , Modeling
Commentary by Dr. Valentin Fuster
2010;():783-792. doi:10.1115/ICONE18-29799.

The paper describes a new human-machine (HMI) interface of the VR-1 nuclear training reactor at the Czech Technical University in Prague. The VR-1 reactor is primarily used for training of university students and future nuclear power plant staff. The new HMI was designed to meet functional, ergonomic and aesthetic requirements. It contains a PC with two monitors. The first alphanumerical monitor presents text messages about the reactor operation and status; next, the operator can enter commands to control the reactor operation. The second graphical monitor provides parameters of reactor operation and shows the course of the reactor power and other parameters. Furthermore, it is able to display the core configuration, perform reactivity calculations, etc. The HMI is also equipped with an alarm annunciator. Due to a high number of foreign students and visitors at the reactor, the Czech and English language versions of the user interface are available. The HMI contains also a History server which provides a very detailed storage and future presentation of the reactor operation. The new HMI improves safety and comfort of the reactor utilization, facilitates experiments and training, and provides better support for foreign visitors.

Topics: Machinery
Commentary by Dr. Valentin Fuster
2010;():793-801. doi:10.1115/ICONE18-29820.

An Adaptive Unbalance Vibration Control (AUVC) method is studied based on the flexible active magnetic bearing-rotor experimental system AMB-PII, which is built up to simulate the helium turbo compressor rotor of the 10MW high temperature gas-cooled reactor with direct cycle (HTR-10GT) in compliance with the dynamic similarity principle. By constructing the required frequency compensation signal automatically through different feedforward approaches, two types of control methods are realized: the “displacement nulling” control which can effectively cancel the rotor vibration amplitude within the system bandwidth, and the “current nulling” method which can let the rotor rotate around its inertia axis and eliminate the disturbances of the currents in the electromagnet windings to significantly attenuate the mechanical vibration. The simulation and experimental results have proved that this kind of AUVC integrating feedforward algorithm can achieve good performance improvements, which will provide good experience and design references for the future application of magnetic bearings in the HTR-10GT project.

Commentary by Dr. Valentin Fuster
2010;():803-807. doi:10.1115/ICONE18-29922.

In the field of system and control, the treating for nonlinear system is always the hot issue in present research. The multi-model method is an important method for solving the nonlinear problem. In traditional solution, the performance of multi-model simulation is not convenient enough. In this paper, a simple, visual and convenient simulation technique is introduced. The toolbox “Simulink” which is derived from the software “Matlab” is used for simulating dynamic process. As a new original attempt, the transfer functions of system models are transformed to module charts. According to the module charts, the multi-model system module is built. The advantage of this technique is that: the coefficients in transfer functions are online controllable. Then the switch and connecting rules between sub-models are introduced. The switch and connecting rules are designed according to the actual case. All of above elements will be designed to module in Simulink workspace. Finally, an illustration is taken. With using the method recommended in this paper, the operation of the simulation is visual, simple and convenient. The technique recommended in this paper can also be suitable for other fields which include nonlinear system.

Commentary by Dr. Valentin Fuster
2010;():809-814. doi:10.1115/ICONE18-29928.

The US-APWR, currently under Design Certification review by the U.S. Nuclear Regulatory Commission, is a four loop evolutionary pressurized water reactor with a four train active safety system applied by Mitsubishi Heavy Industries. The digital Instrumentation and Control (I&C) System and Human Systems Interface (HSI) system are to be applied to the US-APWR. This design is currently being applied to the latest Japanese PWR plant and to nuclear power plant I&C modernization program in Japan. The US-APWR digital I&C and HSI system (HSIS) utilizes computerized systems, including computer-based procedures and alarm prioritization, relying principally on an HSIS with soft controls, console based visual display units (VDUs) and a large, heads up, overview display panel. Conventional hard-wired controls are limited to system level manual actions and a Diverse Actuation System (DAS). The overall design philosophy of the US-APWR is based on the concept that operator performance will be enhanced through the integration of safety and non-safety display and control systems in a robust digital environment. This philosophy is augmented, for diversity, by the application of independent safety soft displays and controls. In addition, non-digital diverse automatic and manual actuation system is introduced. As with all the advanced designs, the digital systems open as many questions as they answer. To address these new questions, for an eight week period during the months of July and August 2008, an extensive verification and validation (V&V) program was completed with the objective of assessing US operators’ performance in this digital design environment. (Robert E. Hall et al., 2008, “US-APWR Human Systems Interface System V&V Results: Impact on Digital I&C Design”, 17th International Conference on Nuclear Engineering, ICONE17-75176) [1] Over this time period, U.S. operating crews were subjected to exercise in Mitsubishi dynamic simulator. To follow up above mentioned V&V activities, additional test during the months of this spring in 2009 has been carried out to resolve human engineering discrepancies (HEDs) induced from the previous evaluation and the participants’ comments and performance. Subjective and objective data were collected on each crew for each scenario and an extensive convergent measures analysis was performed, resulting in the identification of both specific design as well as generic conclusions. This paper discusses the digital HSIS of the US-APWR design, the V&V program data collection and analysis, and the study results related to the ongoing discussion of the impacts of digital systems on human performance, such as workload, navigation, situation awareness, operator training and licensing.

Topics: Design
Commentary by Dr. Valentin Fuster
2010;():815-820. doi:10.1115/ICONE18-29958.

This paper describes a versatile test facility developed by AECL for validation and reliability (V&R) testing of safety-critical software used in the process trip computers for CANDU reactors. It describes the hardware and software aspects of the test facility. The test hardware consists of a test rig with a test computer used for executing the test software and a process trip computer emulator. The test software is comprised of an operating system, a test interpreter, a test oracle, and a man-machine interface. This paper also discusses the application of the test facility in V&R testing of the process trip computer, how test scripts are prepared and automatically run on the test computer, and how test results are automatically generated by the test computer, thus eliminating potential human errors. The test scripts, which contain specific instructions for testing, are text files written in a special AECL test language. An AECL Test Language Interpreter (ATLIN) program interprets the test scripts and translates structured English statements in the test scripts into test actions. The intuitive nature of the special AECL test language, the version controlled test scripts in text format and automatic test logging feature facilitate the preparation of test cases, which are easy to repeat, review and readily modifiable, and production of consistent results. This paper presents the concept of adding a process trip computer emulator for use in preparation of V&R testing. The process trip computer emulator is designed independently from the actual process trip computer but based on the same functional specification as for the process trip computer. The use of the process trip computer emulator allows the test scripts to be exercised before the actual process trip computers are available for V&R testing, thereby, resulting in a significant improvement to the project schedule. The test facility, with the built-in process trip computer emulator, is also a valuable training tool for the V&R staff and plant personnel.

Commentary by Dr. Valentin Fuster
2010;():821-826. doi:10.1115/ICONE18-30055.

The application of Coriolis flowmeters to various industries has been expanding since they were successfully introduced in the market some 30 years ago. This paper specifically reports the development of Coriolis flowmeters for the nuclear industry based on the latest straight-tube technology. An advanced numerical method combining the strength of commercially available simulation packages and in-house theoretical work was used in the development. This includes the initial prediction of flow sensitivity using the fluid-structure interaction theory for straight flow tubes to ensure sufficient measurement accuracy under various process conditions. It also includes further detailed analysis of the entire flowmeter to make sure its natural frequencies away from the seismic frequency range due to nuclear application safety reasons. Furthermore, a new calibration and testing procedure is reported in this paper, which includes the normal calibration condition and also simulates possible process conditions during the nuclear applications (e.g. varying fluid temperature). Additionally, calculations based on the Design by Formula approach according to the governing codes, particularly ASME Boiler and Pressure Vessel Code, Section III, are also reported. These simplified calculations are of significant importance for Coriolis flowmeters to be applied to nuclear applications. Finally, a case study which involved significant collaboration with a well-known safety-related solution provider in the nuclear industry is described, where Coriolis flowmeters were required to provide accurate measurement for the boric acid make-up subsystem in a pressurised water reactor system. Design checks and special considerations for the fabrication and testing according to the rules of ASME Code Section III together with a quality assurance procedure as described in the case study showed that the developed Coriolis flowmeter can meet the specific requirements of nuclear applications in terms of both measurement performance and safety concerns.

Topics: Flowmeters
Commentary by Dr. Valentin Fuster
2010;():827-834. doi:10.1115/ICONE18-30074.

Mitsubishi digital safety Instrumentation and Control (I&C) system has been developed and approved in Japan. The digital I&C system has been applied to many safety and non-safety system applications including full digital I&C system for new plants and digital upgrading for operating plants in Japanese Pressurizer Water Reactor (PWR) plants. The digital I&C system ensures defense-in-depth and diversity for plant safety and control, and this feature also provides countermeasures against software common cause failures. Based on this proven technology, the digital I&C system will also be applied for the US-APWR in the U.S. plant. The US-APWR is one of the candidate reactor of future nuclear power plants in U.S., which has been developed by Mitsubishi Heavy Industries, Ltd. (MHI) modifying Japanese Advanced Pressurizer Water Reactor (APWR) design to comply with U.S codes and standards. The I&C system of the US-APWR also conforms to the U.S. regulatory requirements and industry guidelines. The safety I&C system design and digital platform for the US-APWR are summarized into topical reports and are currently reviewing by U.S. Nuclear Regulatory Commission (NRC). The I&C system includes Human System Interface System (HSIS), Protection and Safety Monitoring System (PSMS), Plant Control and Monitoring System (PCMS) and Diverse Actuation System (DAS). The paper describes our digital I&C design features and application of the digital I&C system to new plants and digital upgrading for operating PWR plants.

Commentary by Dr. Valentin Fuster
2010;():835-840. doi:10.1115/ICONE18-30095.

This paper presents the results obtained from the IBE-CNC/DAQ-090827 project, conducted by the company “Titania Servicios Tecnológicos, S.L.” in collaboration with the “Instituto de Seguridad Industrial, Radiofísica y Medioambiental” (ISIRYM), in the “Universidad Politécnica de Valencia”, for the company “Iberdrola Generación S.A”. The objective is the acquisition of the pressure sensor signal and the measurement at points C85 and N32 from the cabin of the Turbine Control System in Cofrentes Nuclear Power Plant. With the study of previous data, one can obtain the Bode plot of the crossed signals as requested in the technical specification IM 0191 I. Frequency response (i.e. how the system varies its gain and offset depending on the frequency) defines the dynamics.

Commentary by Dr. Valentin Fuster
2010;():841-844. doi:10.1115/ICONE18-30099.

CPR1000 plant operation is based on a centralized and computerized control mean (the KIC system), that provides all the means needed by the shift team to operate the plant in any situation. There are two main consequences regarding the definition of the KIC configuration. 1) The KIC composition shall be defined so that they are sufficient for the shift team to finish their tasks. It is linked to “Human Factors” concerns as well as organization concerns. 2) The probability of KIC OWP (Operator Work Place) unavailable shall be taken into account in order to figure out the relationship between the fault leading to KIC OWP unavailable and different KIC degraded situation. The technology and architecture choices fix limits to the flexibility and robustness that the KIC can offer regarding its own degraded situations. Therefore, in order to keep a low frequency of use of the Back-Up Panel, it might be needed to adapt and review the operating organization and needs to stay on the KIC by raising the discomfort of operation for the shift team. Nevertheless, such an approach is limited as reducing the discomfort of operation as obviously a price in term of safety notably. Therefore, the KIC degraded situation shall be studied in order to identify the criteria beyond which the safety operation of the plant cannot be ensured anymore from the KIC. After studying the features of the fault leading to KIC OWP key component unavailable, the data from DCS supplier, the data from reference NPP (Nuclear Power Plant) and the statistic during commissioning, the MTBF (Mean Time Between Failures) for each OWP has been calculated. Applying Markov Processes based on the studying above, CPR1000 operation principles, features of SOP (state oriented procedures to deal with accident situation) and PSA, CNPEC suggest the method to calculate KIC minimum configuration for reaching the best balance to maximize the availability of the KIC, to minimize frequency of switch to the Back-Up Panel (conventional control means) and to increase the plant availability (as the switch to the Back-Up Panel can lead to the fallback of the unit).

Commentary by Dr. Valentin Fuster
2010;():845-850. doi:10.1115/ICONE18-30183.

The detailed design of the AP1000 power plant control rooms and human system interface (HSI) is ongoing. Task analysis (TA) is conducted and incorporated into the design of the control rooms and HSI for AP1000 to ensure that the indications and controls provided by the HSI resources enable the operator to perform the tasks required to control, monitor and maintain the AP1000 plant safely and efficiently. The major focus of the AP1000 TA is the main control room (MCR). However, TA is also conducted for the technical support center (TSC), the emergency operations facility (EOF) and maintenance, test, inspection and surveillance (MTIS) tasks. The human factors team has applied various TA methods in their analyses, including function-based task analysis (FBTA), operational sequence analysis (OSA) and hierarchical task analysis (HTA). The FBTA is based on an existing and established functional decomposition (goal-means analysis) for normal and emergency operations of the AP1000. At the top of the FBTA are two goals: 1) generate electricity and 2) prevent radiation release. These goals are decomposed to increasing levels of detail showing sub-goals or functions that satisfy the goal in the level above. The OSA is conducted for a representative set of operational and maintenance tasks. Tasks are selected to represent the full range of operating modes and the full range of activities in the AP1000 emergency response guidelines. Operator and MTIS tasks that are identified as risk-important or are associated with risk-significant systems, structures, components (SSCs) are also selected. The OSA focuses on operational requirements or task demands in terms of operator and maintainer actions or processes necessary to complete the tasks. In addition, HTA is conducted for tasks in which the data and displays available in the MCR may be utilized in the TSC and/or EOF. Four tasks were identified and analyzed in support of TSC/EOF design. The results of the HTA for the TSC/EOF are arranged by goals, tasks, and sub-tasks, and identify the information requirements to support decision making. The results of the TAs and generated recommendations are incorporated into the AP1000 HSI design as inputs at an early stage in the design process in a timely manner. The results of TAs also provide inputs to the development of procedures, staffing, training, and communication requirements.

Commentary by Dr. Valentin Fuster
2010;():851-853. doi:10.1115/ICONE18-30186.

The thesis researches the Safety Information and Control System (SICS) design principle and introduce engineering application in CPR1000 nuclear power station in China. The SICS provides sufficient control and monitoring means to bring and maintain the plant in a safe state as a backup of Main computerized Control Mean (MCM), in any plant conditions that are probable during a planed or unplanned unavailability of the MCM. The successful engineering application of SICS in different digital I&C system platform are introduced in the paper. The thesis gives the research conclusion for new general SICS of digital I&C system.

Commentary by Dr. Valentin Fuster
2010;():855-860. doi:10.1115/ICONE18-30236.

This paper describes the power coordinated control system in PWR NPP. The emphasis of this paper is to describe the key parameters selection, several coordinated control solution and operation mode selection. The some factors should be considered during control system design are also introduced in this paper.

Commentary by Dr. Valentin Fuster
2010;():861-865. doi:10.1115/ICONE18-30237.

According to the study of series of related standards, the defense-in-depth and diversity have become the basic requirements for digital I&C and should be strictly followed when carrying out the overall scheme design of Digital Control System, in which the risk of Common Cause Failure has to be prevented or coped with. This paper analyzes the requirements according to the diversity principle and illustrates the typical solution of the diversified protection base on the CPR1000 NPP. Some concepts will be helpful to the design of DCS.

Commentary by Dr. Valentin Fuster
2010;():867-871. doi:10.1115/ICONE18-30239.

An instrument based on a small and inexpensive silicon diode was developed and used to measure, for the first time, the gamma radiation exposure rate within a start-up instrumentation guide tube in a shut down CANDU® reactor undergoing refurbishment. The shape of the measured profile agreed with the expectation; however, the maximum measured exposure rate was about four times higher than the calculation. Two adjuster rods located above the core and adjacent to the guide tube, could be clearly identified from the features of the radiation profile measurement. These measurements provide confirmatory information for the creation of safe and effective radiation work plans and selection of appropriate in-core instrumentation. This technology can be used to accurately measure the positions of irradiated reactor components, such as adjuster rods, pressure tubes, calandria tubes, reactivity control absorbers and other sources of gamma radiation inside a shut down reactor.

Topics: Gamma rays
Commentary by Dr. Valentin Fuster
2010;():873-876. doi:10.1115/ICONE18-30278.

This paper discusses how reactor plant instrumentation and control (I&C) diversity implementation is achieved and impacted with evolving U.S. and worldwide diversity regulatory guidelines and requirements. With increased reliance on digital-based I&C architectures and risk-informed insight methods, diversity regulatory guidelines and requirements are still evolving. In some cases, the regulators are prescribing different methodologies to achieve the common goal of an adequate level of defense in depth through diversity. This paper analyzes differing diversity methodologies to compare and contrast rules as applied to actual diversity implementation on I&C systems.

Commentary by Dr. Valentin Fuster
2010;():877-881. doi:10.1115/ICONE18-30317.

Class IE electric cables can generally withstand 40 years of degradation from predicted operational environments and still perform safety-related functions even during and after accidents. These cables degrade through deterioration of insulation material, possibly leading to mechanical and electrical failure after decades of operation. The elongation at break is usually used as the critical parameter. However, elongation testing is destructive and requires relatively large specimens, making it undesirable for analyzing installed cables. In this paper, indenter modulus and break-elongation test for condition assessment of NPP cables were performed after accelerated ageing under heating and radiation. The test results showed that the relation between break-elongation and indenter modulus, and concluded that Indenter modulus can be effectively used to condition assessment of NPP cable ageing degradation.

Topics: Cables , Elongation
Commentary by Dr. Valentin Fuster
2010;():883-892. doi:10.1115/ICONE18-30320.

Digital control and safety plus the complete functional and physical separation between control and safety and also between the safety systems have been key long standing principles of CANDU® nuclear reactor technology. This paper presents a historical evolution of these principles that make CANDU reactors one of the safest technologies in the world today. The original Generation II CANDU 6 reactors started with complete separation of control from safety and the division of safety systems into two groups having strong physical separation such as opposite sides of the reactor or reactor building. Within each group a more moderate distance separation was employed. With the advent of distributed computer technology for control and display functions, key processing equipment is now moved out remote from the control rooms and distributed into channelized field equipment rooms around the reactor building as in the Four-Quadrant concept for ACR-1000™. This new approach is immune to total unavailability of any control room or equipment room due to events such as fire with minimal impact to any of the safety systems regardless of their grouping. In addition to physical separation, appropriate functional partitioning, design rules to avoid communication cross links, and diversity principles are applied to computer based I&C systems as defences against common cause faults.