Commercial Power Reactors Research Articles

Steel developments for improved fatigue resistance in suspension springs: Smart, E., Cook, W. T. and Ellis, A. Ironmaking and Steelmaking 1992 19 (4) 285–289

The properties of carbon and low alloy steels currently used in the USA for commercial nuclear power reactor pressure vessel construction are examined in reference to higher strength-higher alloy steels having potential for advanced reactor service. Emphasis is placed upon the characteristics of selected vessel steels which are critical to the intended service function and the influence of fabrication and the service environment in changing important mechanical properties. Recent developments for assessing these properties after simulated nuclear service are reviewed along with techniques for minimizing the effects of radiation. The review places much emphasis upon improved properties for steels which show strong potential for future nuclear structural application. Engineering considerations are cited only to highlight the descriptive review of pressure vessel steels, current and future, and their capacity for meeting the critical functions required of a reactor primary containment vessel.

On the utility of tokamaks for energy

At the 14th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, Germany, several senior experts on magnetic fusion development shared with me their view that finally, the world fusion effort has accumulated enough understanding of the requirements and conditions for ignited long-pulse burn, that a tokamak experiment to achieve that objective could be built and expected to work with high confidence. Although this is not yet a unanimous view of all the IAEA participants, I do agree with this majority view. As a reflection of this growing confidence, the governments of the United States, the European Community, Russia and Japan have recently commissioned the Engineering Design Activity (EDA) phase of the International Thermonuclear Experimental Reactor (ITER). The technical vision of ITER presented at that conference by the new ITER director, Paul-Henri Rebut, was widely accepted as a most practical and workable approach to achieving the above objective, assuming a successful demonstration of the proposed gas-target divertor during the EDA R&D phase. Having pioneered an early version of the gas-target divertor [l], I also agree with this assessment of the technical viability of Rebut’s proposed design. At the end of his ITER presentation, however, Rebut touched on an issue of much lower confidence among the fusion community, namely, that ITER would demonstrate “a reasonable cost per unit of fusion power”. Estimating myself that the magnet set (TF + PF) of his proposed design would store about 80 GJ, would have a cold mass exceeding 13000 tons, including magnet support structure, and would alone cost more than 5 B $ at 4OO$/kg (for 13 T max, Nb,Sn), I take a charitable interpretation of Rebut’s remark to mean that ITER demonstrate a reasonable unit cost, not by itself alone, but viewed as a significant step toward a practical power reactor. This then shifts the confidence issue on economics to a judgment on how much improvement in unit cost can be reasonable expected in tokamak reactors beyond ITER. Improvement beyond ITER was an important motivation behind the recent recommendation by the US Secretary’s Energy Advisory Board (SEAB) to proceed with the tokamak Physics Experiment (TPX) at the Princeton Plasma Physics Laboratory. This prompts the question as to how much improvement in unit cost would be needed from the present embodiment of ITER to an economically-competitive power reactor, assuming a 1 GW, power reactor would be at least no larger in size or mass than ITER. I would estimate the present expectation of ITER’s unit cost, based on a “median” performance with operation at prudent margins below accepted limits, to be = 10 B $/l GW,,, or, equivalently, 25$/W,, as if hot blankets and a 40% efficient conversion system were included. For a “competitive” commercial power reactor, I would take 2.5$/W, (based on total capital as in the ITER number). Thus, by this figuring, I estimate than an improvement factor of about 10 is needed (I wouldn’t quibble between 7 and 14).

Systems Engineering for Decommissioning the Japan Power Demonstration Reactor.

The Japan Power Demonstration Reactor (JPDR) decommissioning program is in progress : new technology for reactor decommissioning is being developed and a database is being established. Various data pertaining to dismantling of the JPDR have been accumulated in the database. The data collected in the program are being used for (1) managing ongoing dismantling activities, (2) verifying the code system for management of reactor decommissioning (COSMARD) developed for the forecasting of management information, and (3) planning future decommissioning of commercial nuclear power reactors. COSMARD and the database are expected to be particularly useful in the last of these three.

Measurements of 129I in a nuclear power reactor by accelerator mass spectrometry

The method of accelerator mass spectrometry is applied to measurements of concentrations of the long-lived volatile fission product 129I ( t 1 2 = 1.6 × 10 7 a) in the primary heat transport (PHT) system and irradiated fuel bay (IFB) of a commercial nuclear power reactor. Concentrations of (8.1 ± 1.1) × 10 12 129I atoms per gram and between 10 11 and 10 13 129I atoms/L were measured in an ion-exchange resin and in the heavy water (PHT) system, respectively. The corresponding value for the IFB system water samples was 5–6 × 10 12 129I atoms/L. These measurements, based on very small size samples (of the order of 0.5 mL) emphasize the advantage of the present method where a simple chemical process for the iodine extraction enables a direct measurement of the 129I atom content of the samples. In view of the physical and chemical properties of iodine and their important implications in the environment, the method is likely to find even wider applications for the monitoring of 129I and the study of long-range nuclear waste storage.

Solar and supernova neutrino physics with Sudbury Neutrino Observatory.

The salient features of solar and supernova neutrino physics at the Sudbury Neutrino Observatory are examined. The possibility of detecting a solar-antineutrino flux, predicted as a result of resonant spinflavor conversion, is emphasized. The antineutrino flux at Sudbury due to the commercial power reactors is calculated. Simplified, analytical expressions for the expected galactic supernova count rates are given.

Recommendations for effective loose parts monitoring

Loose parts monitoring systems are used to detect structure borne sound that can indicate the presence of impacting metal parts inside the primary coolant system of commercial nuclear power reactors. Experience with these systems has not been uniform, with cases including high false alarm rates, failure to detect loose parts, and limited capability for interpreting alarms and signals. Utilities have become increasingly aware of benefits that can be obtained from loose parts monitoring as a result of events that damage steam generator tubes, require fuel and reactor internals cleaning, and cause fuel pin failures. A research project, sponsored by the Electric Power Research Institute, used metal impact theory and experimental data to develop a quantitative description of loose part signals as a function of impact, signal transmission, and detection properties. These signal properties were used to develop recommendations for the number, location, and attachment of sensors, and methods for signal processing, recording, impact response calibration, system surveillance, and signal interpretation. Technical considerations and experience link these systems to combined applications for component vibration and rotating equipment monitoring.

ESTIMATING THE COSTS FOR JAPAN’S JPDR PROJECT

There has been an international flavor of cooperation in the commercial development of nuclear power. This cooperation is most strongly established with the development of strategies for decommissioning. Many countries are beginning this phase at about the same time; Japan, Canada, Germany, and the United States all are involved in dismantlement projects. This chapter, along with the next four chapters, addresses various aspects of the methodological approaches being developed to estimate decommissioning costs. Each country has adopted a different strategy for reactor decommissioning, taking its political and technical situations into consideration. Japan also has its own strategy of reactor decommissioning, reflecting the geographic and economic positions peculiar to Japan. The Japan Power Demonstration Reactor (JPDR) decommissioning program is in the process of establishing a decommissioning data base and a cost estimation methodology, as well as developing new technology for reactor decommissioning. Various information about the JPDR dismantling has been accumulated in the decommissioning data base, which will be used for: (1) planning future decommissioning of commercial nuclear power reactors; (2) verifying the developed code system for management of reactor decommissioning; and (3) managing the ongoing JPDR dismantling. The computer code system developed in this program is expected to contribute to studying cost estimation and the optimization of decommissioning plans for commercial nuclear power reactors.

Divertor Engineering for the ARIES-I Reactor†

The ARIES tokamak-reactor design study is a multi-institutional project investigating several different visions of the tokamak as a commercial power reactor. The ARIES-I reactor design incorporates modest extrapolation of existing physics with advanced technology in the fusion power core (FPC) design. Use of aggressive technology such as high-field magnets and low-activation silicon-carbide (SiC) composites help to make ARIES-I an attractive reactor with excellent safety characteristics. The ARIES-I reactor uses a double-null poloidal-field (PF) divertor for particle exhaust and impurity control. Control of the edge-plasma density and temperatures has reduced the energy of the incident particles at the divertor to ∼20 eV, which is below the self-sputtering threshold for the tungsten target surface material. The target is constructed of SiC composite tubes with a 2-mm-thick plasma-sprayed coating of tungsten on the plasma-facing side and a 0.5-mm chemical-vapor deposited (CVD) coating of SiC on the back. The divertor is cooled by helium at lOMPa with inlet/outlet temperatures of 350°/650°C. In removing the divertor surface heat flux of 4.5 MW/m2, a design safety factor of 1.8 is achieved. The divertor has a waste disposal rating of 0.10 (see text for definition), thus allowing Class-C shallow land burial, and a site boundary dose of 11.2 rem during an accidental release. Isotopic tailoring of the tungsten and target replacement every two years is necessary to achieve these safety characteristics.

24 Tesla superconducting toroidal field magnet concept for a commercial Tokamak reactor

A 24-T superconducting toroidal field magnet concept for a commercial fusion power reactor is discussed, and a magnet development path is outlined. Superconducting and structural materials options are discussed within the framework of a 20-yr development time. Nb/sub 3/Sn may be capable of operating at up to 18-20 T. For B>20 T, Nb/sub 3/Al, Nb/sub 3/(Al,Ge) or ceramic high-T/sub c/ superconductors are required. Enhanced strength and stiffness and reduced conductor bending stress are obtained with carbon fiber reinforced steel (Incoloy 908) as the primary structural material. Composite behavior is analyzed using a two-dimensional orthotropic failure criterion. The size of the magnet is minimized subject to electromagnetic, mechanical, and practical constraints.

Advances in high fluence solid state track recorder neutron dosimetry

Neutron dosimeters based on measurements of neutron-induced fission rates with solid state track recorders (SSTRs) provide a versatile and accurate method for reactor dosimetry applications. The high fluence upper limit of the range of the SSTR neutron dosimetry method is determined by track counting limitations at very high track densities. The recent development of techniques for the preparation and calibration of ultra low-mass fissionable deposits has extended the high-fluence limit to about 5×10 18 neutrons/cm 2 . To date, more than 900 SSTR neutron dosimeters have been exposed in high neutron fluence applications at commercial nuclear power reactors. At the same time, parallel efforts have been carried out to further extend the range of the SSTR neutron dosimetry technique, to improve the precision of the present methods, and to verify the high accuracy of the method through controlled exposures in standard neutron fields. The present status of high fluence SSTR neutron dosimetry techniques as well as recent improvements in the method will be discussed.

JPDR decommissioning program — plan and experience

The Japan Power Demonstration Reactor(JPDR) decommissioning program has been carried out to develop the necessary techniques for the decommissioning of commercial power reactors in the future by the Japan Atomic Energy Research Institute(JAERI). The JPDR decommissioning program consists of two major phases. Phase I started in 1981 aiming at the development of a set of techniques necessary for reactor dismantling. Phase II, actual dismantling of JPDR using techniques developed in phase I, started in December 1986. The dismantling work began in the containment and the dump condenser buildings. After removing the peripheral components around the reactor, dismantling of highly activated JPDR internals began during the period of January to March 1988 using the underwater plasma arc cutting system operated by a robotic manipulator. The effectiveness of various dismantling techniques developed in the JPDR decommissioning program has been verified through the actual dismantling activities.

Nuclear Waste: the challenge is global

The difficulties in disposing of many millions of tons of nuclear waste, comprising radioactive elements with half-lives ranging from microseconds to eons, are discussed, stressing the global nature of the issue. The IAEA classification of nuclear waste into high, intermediate, and low-level forms is described. The classification is based on several factors, including the waste's source and temperature, and half-life: the greater the number of long-lived radioisotopes the waste contains, the higher the level of waste. Although, overall, wastes from commercial power reactors outweigh those from military ones, military programs produce a huge excess of the high-level type. Methods of reprocessing high-level wastes, the solution most favored in Europe, are examined. Opposition to permanent underground disposal sites, the principal alternative to reprocessing, is discussed. Progress in disposing of low-level wastes is described. The technology of vitrification of high-level wastes and of nuclear fuel reprocessing is examined. >

Economic Assessment of Tokamak Fusion Reactor Plants

As the R&D of nuclear fusion technology makes steady progress, it becomes important to examine with what perspective and toward what direction further steps should be taken. This paper discusses the prospect of commercial nuclear fusion power reactors principally from the viewpoint of electric utilities, namely its potential users. Based upon conceptual designs of pure fusion reactors and fusion fission hybrid reactors, generation costs (cost of electricity) of the reactors are evaluated. Two different approaches are employed in the design because much uncertainty still remains in the future characteristics of fusion power reactors. The generation costs for the above two approaches show a fair similarity. The evaluated costs of the hybrid reactors are significantly lower than those of the pure fusion reactors.

Materials Requirements for Experimental Fusion Reactors

The next major step in the development of fusion power involves the design, construction, and operation of an experimental power reactor. The general objective of this type of device is to demonstrate key plasma physics, materials, and engineering performance under conditions that approach those considered necessary to establish the technical feasibility of fusion power. This device is generally perceived to be a tokamak configuration that operates on the deuterium-tritium (D-T) fuel cycle, which would produce several hundred megawatts of fusion power. Although a number of conceptual design studies on this type of device have been conducted, the current effort is focused on the International Thermonuclear Experimental Reactor (ITER) activity, a joint undertaking of the European Community (EC), Japan, the United States, and the Soviet Union, conducted under the auspices of the International Atomic Energy Agency (IAEA).The development of fusion power as an energy source depends to a large extent on the proper selection of materials for the various components. these materials will be exposed to a wide variety of conditions such as plasma particle and neutron radiation, high thermal fluxes, large thermal and mechanical stress, various chemical environments, and high magnetic fields. In order to achieve the high performance desired for a commercial power reactor, considerable materials development will be required. Because of the relatively near term time scale projected for ITER and the limited objectives, more conventional materials with more extensive data bases will be used where possible. Even so, a substantial materials development program is required to support a test reactor like ITER.

Numerical studies of deuterium-tritium ignition in impact-fusion targets

A numerical one-dimensional solution of the Euler equations for an imploding spherical tungsten shell with internal deuterium-tritium gas is applied to study impact-fusion dynamics with parameters of fusion-reactor relevance. Thermal conduction and radiative energy loss by the plasma are taken into account, as is heating by fusion-generated alpha particles. A variety of target sizes and impact velocities are examined, and scaling laws for fusion yields are deduced which define possible parameters for conceptual commercial impact-fusion power reactors. It is found that shell energies and velocities of about 30 MJ and 110 km/s would be satisfactory. A potential commercial impact-fusion reactor based on such parameters is discussed.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Plutonium Recycling and the Problem of Nuclear Proliferation

A typical 1 gigawatt LWR, the dominant commercial power reactor type today, operating at 70 per cent capacity factor, generates approximately 250kg of plutonium annually. This plutonium, which is produced in the reactor through neutron capture by U-238, is then discharged from the reactor along with the other constituents of the spent fuel. About 0.7 per cent of the plutonium, or 175kg, is fissile (odd) isotopes of plutonium. As long as the plutonium discharged from the reactor is left intermixed with the highly radioactive fission products also contained in the spent fuel, it cannot readily be used for power or for weapons.

Pressure vessel surveillance at commercial light water reactors using solid state track recorder neutron dosimetry

Abstract The accumulation of neutron dose by the pressure vessel of a light water reactor results in radiation damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, the neutron dose must be monitored. Solid state track recorder neutron dosimeters have been developed for pressure vessel surveillance dosimetry at commercial power reactors. To date a total of more than 200 neutron dosimeters from seven different power reactor dosimetry exposures have been analyzed, and more than 150 additional dosimeters are currently deployed. The techniques for carrying out these measurements and thier results will be discussed.

Environmental Effects of Fusion Power Plants. Part I: Effluents Other than Tritium

Radioactive and nonradioactive effluents are released routinely during normal operation of near-term commercial fusion power reactors. Nonradioactive effluents are essentially the same as those released at conventional steam-electric power plants. Radioactive effluents consist of activated corrosion products and tritium. Most radioactive releases originate from liquid-waste processing systems and from ventilation systems of various buildings where radioactivity may become airborne. These effluents have some potential for environmental impact; however, the significance of the impact depends in part on the concentration and release rate of the effluent. The type of reactor design (e.g., tokamak, mirror, etc.) has minimal influence on activation product releases. Activation products released are influenced primarily by the materials chosen for structural components, and the quantities released are influenced primarily by the coolant choice.

AGENDA SETTING AND “NONDECISIONMAKING”: DECOMMISSIONING NUCLEAR GENERATING STATIONS

The final point in the nuclear fuel cycle has always been the eventual retirement, decommissioning, of the 81 commercial nuclear power reactors presently in existence. This eventual retirement has been thought to be an issue of the future, because the large plants, built in the 1960s and 1970s. were assumed to have an expected operating life of 30 to 40 years. However, the reality is that several commercial reactors already have been decommissioned, and numerous others will soon reach maturity. This article examines decommissioning of nuclear power plants from a public policy–rather than a technical–perspective. A number of questions are addressed concerning the policy implications associated with decommissioning. The findings of the study are based on survey data from the utilities, an examination of NRC documents, interviews with NRC staff scientists, and site visits to several decommissioned plants.

Physical and chemical characteristics of lead-iron phosphate nuclear waste glasses

Experimental determinations of the properties of lead-iron phosphate glasses pertinent to their application to the problem of the permanent disposal of high-level nuclear wastes have been carried out. These investigations included studies of the composition and physical properties of nuclear waste glasses (NWG), as well as the effect of preparation conditions. Lead-iron phosphate nuclear waste glasses were prepared by dissolving simulated US defense wastes or simulated commercial power reactor wastes in molten lead-iron phosphate melts at temperatures between 900 and 1050°C. The measured physical and chemical properties of the nuclear waste glasses formed by cooling these melts and annealing included the following: (1) aqueous corrosion resistance as a function of the solution pH, solution temperature, and glass composition, (2) glass density, (3) thermal expansion coefficient, (4) glass transition temperature and softening point, (5) heat capacity, (6) critical rate (7) temperature for the maximum crystallization rate, (8) relative solubility of waste oxides in the glass melt, (9) reactions between the molten glass and the melting crucible (Pt, ZrO 2, Al 2O 3), and (10) studies of possible metal cannister materials. Experimental results for the lead-iron phosphate NWG are compared to available data for borosilicate NWG. Relative to borosilicate NWG, the lead-iron phosphate glasses have several distinct advantages which include a much lower aqueous corrosion rate, a lower preparation temperature, and the ability to immobilize many types of commercial and defense-related high-level radioactive wastes.