Current Light Water Reactors Research Articles

Material Matters

The following article is based on a talk for Symposium X presented by Wolfgang Stoll, Chief Scientific Advisor and Consultant in Siemens, Germany, at the 1996 MRS Fall Meeting.Since 1941 when Glenn Seaborg first isolated plutonium in milligram quantities, the total amount converted through neutron capture in U-238 has increased worldwide to about 1,200 tons and continues to grow about 70 tons/year. What was fissioned in situ in operating nuclear power stations is roughly equivalent to 5 billion tons of black coal, while the fission energy contained in those 1,200 tons unloaded in spent fuel is equivalent to another 2 billion tons of coal. About 260 of these 1,200 tons are ready to release their energy in about 4 kg-portions each in microseconds which is equivalent to 10,000 tons of coal. Most people believe this release of energy poses a major threat of the worldwide arsenal of weapons of mass destruction (WMD). The about 20-fold overkill stored in worldwide WMD is considered superfluous after the crumbling of the Soviet Union. Options are sought to dispose of this surplus in a safe, speedy, and controllable manner. While for highly enriched uranium (HEU) (the other nuclear weapons material) dilution into low-enriched uranium and utilization in current light water reactors (LWR) poses market adaptation problems only, and while the worldwide consensus on the elimination of chemical and biological WMD is still in an initial phase, the decision of both the United States (US) and the former Soviet Union (FSU) to remove most of the plutonium out of weapons looks as if it was a firm political decision.

Core and plant design of the power reactor cooled and moderated by supercritical light water with single tube water rods

A reactor cooled and moderated by supercritical light water with single tube water rods is designed. The plant system is a once-through direct cycle; the whole coolant which flows once through the core is fed to the turbine. This reactor is much simpler than the current light water reactors LWRs, which enhances its economy. The average outlet coolant temperature should be evaluated considering the mixing of hot and cold coolant from the fuel assemblies. To evaluate this, nuclear and thermal hydraulic analysis of the whole core is carried out considering the effect of coolant density feedback to the radial power distribution. The flow rate in the outer region is reduced by the orifice to maximize the outlet coolant temperature. The average outlet temperature is 397°C, which is only 12°C lower than that of the hottest channel since the density feedback flattens the radial power distribution. Axial and local peaking factors are also evaluated. A steam cycle including reheaters, turbines and feedwater heaters is designed for estimating the thermal efficiency. The reactor with orifices has 40.7% thermal efficiency and 1013 MW electric power. The electric power can be larger in the core without orifices and channel boxes since the channel boxes can be replaced with additional fuel rods. The decrease in the average outlet coolant temperature is only 3°C from the core with the orifices.

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions.

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.

Effects of irradiation on intergranular stress corrosion cracking

Intergranular stress corrosion cracking (IASCC) is a pervasive and generic problem in current light water reactor and advanced reactor designs that can lead to widespread component failure. IASCC is believed to be due to either to changes in the grain boundary composition, the microstructure or the water chemistry and corrosion potential. Of greatest interest are the changes in composition and microstructure since IASCC exhibits a well-defined, although not invariant, dose threshold. Changes in grain boundary composition are a result of radiation-induced segregation (RIS) and result in enrichment of nickel, depletion of chromium as well as changes in the impurity element compositions at the grain boundary. Although the basic theory of RIS is believed to be understood, quantitative descriptions of observed changes are not yet possible and hinder the correlation between RIS and IASCC. Changes in the microstructure are intimately linked to the strength and ductility of the irradiated alloy and strong correlations between IASCC and irradiated yield strength have been found. However, a fundamental understanding of the deformation mechanisms and the way in which deformation is coupled to IG cracking in alloys irradiated under LWR conditions (250–360°C, 1–5 dpa) is lacking. Finally, although radiation is known to affect IGSCC through changes in water chemistry and corrosion potential, it is not a necessary condition. Overshadowing and slowing progress on this important problem is a lack of well-defined-data from properly irradiated and properly characterized materials, due principally to inherent experimental and financial difficulties. As such, the specific mechanism(s) of IASCC remain unknown.

Large-break loss-of-collant accident analysis of a direct-cycle supercritical-pressure light water reactor

Large-break loss-of-coolant accident (LOCA) was analyzed in the course of the design study concerning direct-cycle supercritical-pressure light water reactor (SCLWR). The advantages of SCLWR are a higher thermal efficiency and simpler reactor system than the current light water reactors (LWRs). A computer code was prepared for the analysis of the blowdown phase from the supercritical pressure. The calculation was connected to the REFLA-TRAC code when the system pressure decreased to around atmospheric pressure. The analyzed accidents are 100, 75, 50 and 25% cold-leg and 100% hot-leg breaks. First, blowdown and heatup phases without an emergency core cooling system (ECCS) were evaluated. A low-pressure coolant injection system (LPCI) was designed to fill the core with water before the cladding (stainless-steel) temperature reached a limit of 1260°C. The LPCI consists of four units, each of which has the capacity 805 kg/s. An automatic depressurization system (ADS) was designed to release the steam generated in the core in the case of cold-leg breaks and to permit operation of LPCI in the case of LOCAs of less than 100% break. For all cases analyzed, the peak cladding temperatures were lower than the limit when the designed ECCS is implemented.

ALMR fuel cycle economics

The U.S. Advanced Liquid Metal Reactor concept, ALMR, is expected to use recycled metal fuel composed of plutonium, uranium and zirconium. Projections of fuel cycle costs for the ALMR indicate that they will be from 5 to 6 mills/kWh (constant 1987 dollars), which is less than or equal to current light water reactor fuel costs. Since the ALMR fuel costs are not sensitive to uranium ore prices, these costs will be more stable than those of future light water reactors. The ALMR also has the potential to fission actinides from both light water and liquid metal reactor fuels. This ability can reduce the time that high-level radioactive waste are hazardous from millions to hundreds of years with about a 1 mill/kWh increase in cost.

Modeling of Fission Product Release and Transport for Severe Fuel Damage Analyses

In this paper best-estimate calculations of realistic source terms are presented that reduce uncertainties in predicting volatile fission product release from the UO{sub 2} fuel over the temperature range from 770 to 2500 K. The proposed method of correlation includes such fuel morphology effects as equiaxed fuel grain growth and fuel/cladding interaction. The method correlates the product of fuel release rate and equiaxed grain size with the inverse fuel temperature to yield a bulk mass transfer correlation. It is found that fewer and slower releases are predicted utilizing the bulk mass transfer correlation than with the steam oxidation model and the U.S. Nuclear Regulatory Commission's NUREG-0956 correlation. Computational modules are developed to perform the thermal-hydraulic and fission product calculations needed to analyze the severe fuel damage tests. The predictions utilizing the bulk mass transfer correlations overall follow the experimental time-release histories during the heatup, power hold, and cooldown phases of the transients. Good agreements are achieved for the integral releases both in timing and in magnitude. The proposed bulk mass transfer correlations can be applied to both current and advanced light water reactor fuels.

Design considerations for severe accident containment performance

Containment designs for current light water reactors are based on conservative design bases which have resulted in structures that are very rugged with substantial margins to failure. For future advanced light water reactors (ALWRs), severe accident concerns must be considered in addition to the traditional design bases when designing containment structures. Although new, prescriptive regulations for severe accident design bases are not necessary, acceptance criteria for the resolution of severe accident concerns must be established in a timely manner to support the development of these ALWRs. This paper suggests an approach to developing such acceptance criteria and identifies the major severe accident technical issues which need to be resolved for future containment designs.

Nuclear Power Technology: A Mandate for Change

Over the past 10 yr, nuclear power in the United States has been in a very difficult period. No new power plant orders have been placed, orders have been cancelled, construction has been delayed, costs have escalated, and concern for safety has impeded development. Yet, the need for nuclear power in the future appears to be growing. For the United States to play a major role in the nuclear industry of the 21st century, the problems of the past must be identified and corrected. This may include the development of “inherently” or “demonstrably” safe reactors that are not based on current light water reactor technologies.

Status of severe accident research

The authors provide an overview of severe-accident research by (1) describing a generalized low-probability severe accident and (2) providing an enumeration of the individual characteristics of current light-water reactors which are likely to influence the progression of the generalized accident. They discuss current issues in source term calibration. Three research programs that are being conducted to establish a database for evaluating these accidents are described.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Pressurized Water Reactor Inherent Core Protection by Primary System Thermohydraulics

Current light water reactors (LWRs) depend for the protection of core integrity on a multitude of active systems and components, such as instrumentation, cables, electronic logics, relays, actuators, etc., and on human judgment. This approach to safety has led to a complex and expensive plant design in which all parts of the plant where these systems are present must be protected against damage due to, e.g., earthquake. It has also failed to persuade the public about the safety of the reactors because of the existing (but very small) probability of multiple failures leading to core meltdown.With the process inherent ultimate safety (PIUS) approach, this dependence on active systems is eliminated. The safety is now no longer a result of their intervention but is built into the thermohydraulics of the primary system itself. The PIUS primary system response to a number of severe anticipated transients without scram (ATWS) is described, as studied by means of a specially developed computer simulation program.The method is shown by which the thermohydraulic self-protection properties of the primary system terminates these ATWS transients, which could have severe consequences in a conventional LWR, with neither the core nor the rest of the plant suffering any damage (beyond the initial failure assumed).This has important economic consequences. The surveillance and control systems used to run the plant and the buildings in which they are housed can be designed as for a fossil plant, since they no longer have the ultimate responsibility for nuclear safety. The ensuing design simplification pays for the more expensive pressure vessel and primary system. Inherent safety is obtained as a bonus.

Authors

Authors

Decay Rate and Decay Heat Data from Fission Products

Decay rate data of mixed fission products are presented from thermal neutron fission of 233U, 235U, and 239Pu and from fast fission of 238U and 232Th in the absence of neutron absorption by the fission products. The effect of neutron absorption by the fission products on decay rate is shown to be relatively small for conditions that exceed current light water reactor conditions. Decay heat data from fission products of 233U and 232Th in the absence of neutron absorption are provided which supplement similar data for 235U, 238U, and 239Pu given in the new American National Standards Institute/American Nuclear Society 5.1 Standard.

Further Comments on Hazards of Nuclear Power and the Choice of Alternatives

This paper continues an earlier discussion of the hazards of nuclear fission, and of the choice of alternatives (Edsall, 1974), in Environmental Conservation, 1(1), pp. 21–30. Important changes have recently occurred in energy policy in the United States. The Energy Research and Development Administration (ERDA), whose mission is to promote the development of all promising forms of energy production, has replaced the Atomic Energy Commission (AEC). A Nuclear Regulatory Commission, independent of ERDA, has been set up. The proposed ERDA budget, however, is largely dominated by the very heavy costs of research and development for the liquidmetal fast breeder reactor (LMFBR). The AEC, in one of its last acts, produced a voluminous environmental impact statement on the LMFBR. A preliminary version of this statement (spring 1974) was sharply and widely criticized as inadequate, and as highly biased in favour of the breeder. The revised statement is now under critical review (March 1975). The AEC has also produced a comprehensive report (the Rasmussen report) on the safety of the current light-water reactors in use in the U.S.A., with the conclusion that a major accident in a nuclear power-plant is so unlikely as to be almost negligible by comparison with other hazards of daily life. It does not consider the hazards of radioactive waste disposal, or of theft and sabotage. Critics have pointed to what they consider grave flaws in the reasoning of the Rasmussen report, so the matter is still highly controversial.