Carbonactör

Edit

Bioreactor, a device which controls a biologically active environment.

Chemical reactor, a device for containing and controlling a chemical reaction

Fusion reactor, a device for containing and controlling a fusion power reaction

An inductor (possessing reactance) in an electrical power grid

A current limiting reactor is used to limit starting current of motors and to protect variable frequency drives

Nuclear reactor, a device for containing and controlling a nuclear reaction

Reactor (software), a physics simulation engine

The reactor pattern, a design pattern used in concurrent programming

In entertainment

Reactor an alternative title for the 1978 Italian film War of the Robots directed by Alfonso Brescia

Re·ac·tor, a 1981 album by Neil Young and Crazy Horse

Reactor (arcade game), an arcade game created by Gottlieb

Reactor, Inc., a defunct interactive entertainment company founded by Mike Saenz

Reactor, a comedy series on Syfy, hosted by David Huntsberger.

The Reactor (show rod), a show car built by Gene Winfield

S. Kumar, A. K. Puniya, M. Puniya et al., "Factors affecting rumen methanogens and methane mitigation strategies," World Journal of Microbiology and Biotechnology, vol. 25, no. 9, pp. 1557–1566, 2009.View at: Publisher Site | Google Scholar

B. M. Buddle, M. Denis, G. T. Attwood et al., "Strategies to reduce methane emissions from farmed ruminants grazing on pasture," The Veterinary Journal. In press.View at: Publisher Site | Google Schola

C. Martin, D. P. Morgavi, and M. Doreau, "Methane mitigation in ruminants: from microbe to the farm scale," Animal, vol. 4, no. 3, pp. 351–365, 2010.View at: Publisher Site | Google Scholar

M. Shibata and F. Terada, "Factors affecting methane production and mitigation in ruminants," Animal Science Journal, vol. 81, no. 1, pp. 2–10, 2010.View at: Publisher Site | Google Scholar

D. A. Lashof and D. R. Ahuja, "Relative contributions of greenhouse gas emissions to global warming," Nature, vol. 344, no. 6266, pp. 529–531, 1990.View at: Publisher Site | Google Scholar

P. Forster, V. Ramaswamy, P. Artaxo et al., "Changes in atmospheric constituents and in radiative forcing," in Climate Change 2007: The Physical Science Basis. Contribution of Working Group I to the Fourth Assessment Report of the Intergovernmental Panel on Climate Change, S. Solomon, D. Qin, M. Manning et al., Eds., Cambridge University Press, Cambridge, UK, 2007.View at: Google Scholar

IPCC, "Intergovernmental panel on climate change," in Climate Change 2001: A Scientific Basis, J. T. Houghton, Y. Ding, and D. J. Griggs, Eds., Cambridge University Press, Cambridge, UK, 2001.View at: Google Scholar

J. L. Ellis, E. Kebreab, N. E. Odongo, B. W. McBride, E. K. Okine, and J. France, "Prediction of methane production from dairy and beef cattle," Journal of Dairy Science, vol. 90, no. 7, pp. 3456–3467, 2007.View at: Publisher Site | Google Scholar

National Research Council, The Scientific Basis for Estimating Air Emissions from Animal Feeding Operations, National Academy Press, Washington, DC, USA, 2002.

A. J. McMichael, J. W. Powles, C. D. Butler, and R. Uauy, "Food, livestock production, energy, climate change, and health," The Lancet, vol. 370, no. 9594, pp. 1253–1263, 2007.View at: Publisher Site | Google Scholar

D. E. Johnson and G. M. Ward, "Estimates of animal methane emissions," Environmental Monitoring and Assessment, vol. 42, no. 1-2, pp. 133–141, 1996.View at: Publisher Site | Google Scholar

R. M. Murray, A. M. Bryant, and R. A. Leng, "Rates of production of methane in the rumen and large intestine of sheep," British Journal of Nutrition, vol. 36, no. 1, pp. 1–14, 1976.View at: Google Scholar

W. E. Balch, G. E. Fox, and L. J. Magrum, "Methanogens: reevaluation of a unique biological group," Microbiological Reviews, vol. 43, no. 2, pp. 260–296, 1979.View at: Google Scholar

K. D. Ashby, T. A. Casey, M. A. Rasmussen, and J. W. Petrich, "Steady-state and time-resolved spectroscopy of F420 extracted from methanogen cells and its utility as a marker for fecal contamination," Journal of Agricultural and Food Chemistry, vol. 49, no. 3, pp. 1123–1127, 2001.View at: Publisher Site | Google Scholar

P. E. Rouviere and R. S. Wolfe, "Novel biochemistry of methanogenesis," Journal of Biological Chemistry, vol. 263, no. 17, pp. 7913–7916, 1988.View at: Google Scholar

P. N. Hobson and C. S. Stewart, The Rumen Microbial Ecosystem, Chapman and Hall, London, UK, 1997.

R. K. Thauer, A. K. Kaster, H. Seedorf, W. Buckel, and R. Hedderich, "Methanogenic archaea: ecologically relevant differences in energy conservation," Nature Reviews Microbiology, vol. 6, no. 8, pp. 579–591, 2008.View at: Publisher Site | Google Scholar

W. Kühn, K. Fiebig, and H. Hippe, "Distribution of cytochromes in methanogenic bacteria," FEMS Microbiology Letters, vol. 20, no. 3, pp. 407–410, 1983.View at: Publisher Site | Google Scholar

J. L. Garcia, B. K. C. Patel, and B. Ollivier, "Taxonomic, phylogenetic, and ecological diversity of methanogenic Archaea," Anaerobe, vol. 6, no. 4, pp. 205–226, 2000.View at: Publisher Site | Google Scholar

H. Zhang, J. K. DiBaise, A. Zuccolo et al., "Human gut microbiota in obesity and after gastric bypass," Proceedings of the National Academy of Sciences of the United States of America, vol. 106, no. 7, pp. 2365–2370, 2009.View at: Publisher Site | Google Scholar

B. Dridi, M. Henry, A. El Khéchine, D. Raoult, and M. Drancourt, "High prevalence

Reactor control

Thomas W. Kerlin, Belle R. Upadhyaya, in Dynamics and Control of Nuclear Reactors, 2019

8.8 The role of stored energy

Power reactors can supply added steam upon demand before the reactor power changes. This is through use of energy stored in reactor fluid and metal components. Fluids boil and metal components cool down to provide energy. For example, stored energy in a typical PWR can provide around 10 full power seconds per psi of pressure drop. Even in a BWR, stored energy can be used temporarily. Even though opening the BWR main steam valve causes pressure reduction, increased boiling, reduced reactivity and reduced power, this can be tolerated if control action inserts reactivity to cancel the temporary reactivity decrease. The steam delivered to the turbine during this temporary episode is provided mostly by energy stored in saturated water.

View chapterPurchase book

Available and advanced nuclear technologies for nuclear power programs

Bilbao Y LeónS. , ... TyobekaB. , in Infrastructure and Methodologies for the Justification of Nuclear Power Programmes, 2012

APR1400

The Advanced Power Reactor 1400 (APR1400), with a rated power of 1400 MWe, is the largest two-loop PWR currently available. The APR1400 is an evolutionary reactor developed in the Republic of Korea, based on the accumulated experience from the design and operation of the 1000 MWe Korean Standard Nuclear Power Plant and from the EPRI URD (EPRI, 1995, 1999). The APR1400 incorporates a number of improvements to meet operators' needs for enhanced safety, performance and economics and to address new licensing requirements such as the mitigation of severe accidents. The APR1400 has a very characteristic configuration, with two large steam generators and four reactor coolant pumps in a 'two hot legs and four cold legs' arrangement. The APR1400 also features fully digital instrumentation and control (I&C), and a main control room designed with full consideration of human factors. The APR1400 incorporates safety systems with both active and passive characteristics, and has also been designed to take advantage of modularization and prefabrication construction techniques to ensure a predictable construction budget and schedule. Two APR1400 units are currently under construction in the Republic of Korea (Shin-Kori 3 and 4), and they are expected to enter commercial operation in 2013–14. The APR1400 has also been selected for the first four units that will be built in the United Arab Emirates.

View chapterPurchase book

Neutron Chain Reactions

Raymond L. Murray, Keith E. Holbert, in Nuclear Energy (Eighth Edition), 2020

16.6 Neutron Flux and Reactor Power

The thermal power developed by a reactor is a quantity of great interest for practical reasons. Reactor power is related to the neutron population and to the mass of fissionable material present. First, let us look at a typical cubic centimeter of the reactor, containing N fuel nuclei, each with cross-section for fission σf at the typical neutron energy of the reactor, corresponding to neutron speed v. Suppose that there are n neutrons in the volume. The rate (fissions/(s cm3)) at which the fission reaction occurs is thus

(16.14)Rf=nvNσf

If each fission produces an energy w, then the power per unit volume is q‴ = wRf. For the whole reactor, of volume V, the rate of production of thermal energy is P = q‴V. If we have an average flux ϕavg = nv and a total number of fuel atoms nF = NV, the total reactor power is seen to be

(16.15)P=ϕavgnFσfw

From Section 6.4, w = 190 MeV/fission or 1/w = 3.29 × 1016 fissions/(MW s).

Thus, we see that the power depends on the product of the number of neutrons and the number of fuel atoms. A high flux is required if the reactor contains a small amount of fuel, and vice versa. All other things equal, a reactor with a large fission cross-section can produce a required power with less fuel than one with small σf. We recall that σf decreases with increasing neutron energy. Thus for given power P, a fast reactor, operating with neutron energies principally in the vicinity of 1 MeV, requires either a much larger flux or a larger fissionable fuel mass than does the thermal reactor, with neutrons of energy less than 0.1 eV.

The power developed by most familiar devices is closely related to fuel consumption. For example, a large car generally has a higher gasoline consumption rate than a small car, and more gasoline is used in operation at high speed than at low speed. In a reactor, it is necessary to add fuel very infrequently because of the very large energy yield per unit mass, and the fuel content remains essentially constant. From Eq. (16.15) relating power, flux, and fuel, we observe that the power can be readily raised or lowered by changing the flux. By manipulation of control rods, the neutron population is allowed to increase or decrease to the proper level.

Power reactors used to generate electricity produce approximately 3000 megawatts of thermal power (MWt) and, with an efficiency of approximately one-third (33%), supply 1000 MW of electrical power (MWe).

Example 16.7

The reactor in a 1000-MWe power plant is composed of 100 MTU (metric tons of uranium) of 3 w/o fuel. Therefore, the number of U-235 fuel atoms is

nF=m235NAM235=100×106g0.036.022×1023atom/mol235g/mol=7.69×1027atoms

To generate the requisite thermal power, the needed thermal flux is

ϕavg=PnFσfw=3000MWt3.29×1016fissions/MWs7.69×1027atom582.6×10−24cm2=2.2×1013n/cm2s

The flux and criticality are separate descriptors of the reactor condition. The difference can be explained by using an automobile as an analogy. The multiplication factor is akin to vehicle acceleration while the flux is comparable to car speed. Hence, a critical reactor can be operated at many flux, or equivalently power, levels.

View chapterPurchase book

Description of boiling water reactors

Alfonso Prieto-Guerrero, Gilberto Espinosa-Paredes, in Linear and Non-Linear Stability Analysis in Boiling Water Reactors, 2019

2.3.3 Neutron flux monitoring and nuclear oscillations

Reactor power is monitored from the source range up through the power operating range by neutron monitoring channels, which are located inside the reactor core. The in-core neutron flux monitor provides maximum sensitivity to control rod movement during the startup period and provides effective monitoring in the intermediate and power ranges (BWR/6, 1975). Power monitoring in the reactor core is performed with three types of monitors that depend on the power level. One of them is the source range monitoring (SRM), the second is the monitoring in the intermediate range (IRM) and the third is the local power range monitor (LPRM). In the source range, the neutron flux is monitored by SRM with fission counters in a range from source level to 109nv (where nv = n/cm2 s). The IRM that corresponds to the intermediate range is considered to be from 108 to 1013nv, where the neutron flux is monitored by a system using the voltage variance method (Lahey & Moody, 1993).

The local fission range monitor (LFRM) is used in the power range in which the neutron flux is monitored by fixed in-core ion chambers, which are arranged in a uniform pattern throughout the core. These chambers cover a range of 1%–125% of rated power. When a control rod is selected, the readings from the adjacent detectors are displayed on the operator control benchboard together with a display of the position of the rod. Detector assemblies each contain four fission chambers. The chambers are uniformly spaced in an axial direction and lie in four horizontal planes.

In a typical BWR, the average power level is registered in the control room by four average power range monitors (APRM). Each APRM represents the bulk power in the core. The LPRMs do provide inputs to the APRM, which is functionally classified as safety related. Each LPRM houses a fission chamber and their associated signal cables. The LPRMs assignments allow each APRM channel to provide a signal proportional to the average neutron flux in the reactor core. The circuit averages only LPRM signals that are operational and the output from the averaging circuit for each APRM channel is routed to the process computer. The output signals from these monitors are displayed in the control room and are also used to operate trips in the reactor protection system (RPS).

Instrumentation inside the core of a BWR is crucial for an estimate of the thermal power and the other aspects related with the operation of the reactor, as well as the reactor protection system (RPS). It is possible to detect BWR oscillations linked to instability via LPRMs; these detectors are located radially and axially within the core vessel, as depicted in Fig. 2.4. Their task is to monitor the local neutron flux of the reactor at a certain locality. The APRM detectors control the emergency shutdown of a BWR (SCRAM) through a reactor protection system (RPS) mechanism that triggers when the detected APRM oscillation exceeds the security threshold. The in-phase (global or core-wide) oscillations can be observed in the APRM detectors and via the RPS and it is possible to SCRAM the reactor if a strong in-phase oscillation is observed (or the operator can also shut down the reactor if necessary). However, the out-of-phase (regional) oscillations cannot be observed in the APRM detectors, because one out-of-phase oscillation with perfect symmetry (a phase shift of 180 degrees between the reactor core zones that participate in the averaging operation via their respective LPRMs) will cancel the LPRMs averaging, which is not observable in most cases by the APRM. Therefore, the out-of-phase oscillations must be studied at a local LPRM level. Events related to diverging power oscillations have happened before in various BWR facilities in the past. Such events encouraged researchers to develop correction techniques to suppress these events. Nonetheless, in spite of the existence of these corrective methods, unstable events continued to occur. Thus, as an answer to these BWR unstable events, several works were developed to study the physical phenomena behind these events. The detection and suppression mechanisms dedicated to mitigating these unstable oscillations need to identify the type of oscillation through LPRM signal monitoring. The development of methods to detect unstable events is of vital importance in terms of reactor security. The main goal of these methods is to provide a stability indicator (estimated via the study of BWR signals) that grants the operator enough time to act accordingly and in such a way that his actions do not involve a SCRAM straight away. The estimated stability indicator must provide as much information as possible regarding BWR unstable dynamics with enough reliability, precision, and predictive capability to give the operator the time needed to act.



Sign in to download full-size image

Fig. 2.4. Core configuration.

View chapterPurchase book

Power Reactor Control Systems

JEFFERY LEWINS PhD (Cantab), PhD (MIT), in Nuclear Reactor Kinetics and Control, 1978

LOAD DEMAND

Power reactor applications to produce work are diverse; from an extreme, say, of a nuclear marine propulsion plant as an independent unit to a closely integrated central electricity supply (utility). For the former example, load demand arises essentially in the mind of the captain, and control will be required to maintain a steady output at the shaft as wanted and to vary this on demand from the bridge.

The closely integrated utility is a more complicated case. Electrical load commonly shows a number of variations in which only general trends can be anticipated; the trends are overlayed with fluctuations. General trends include the annual cycle—peak loading in winter for heat, light and power as well as (particularly in the United States) a second peak for air conditioning in the summer—and the diurnal cycle—day and evening peaks affected by the time of year, with lows in the early hours of the morning. The size of interconnected utilities make international time zones a part of the demand structure. Random fluctuations can be caused, for example, by the startup or shutdown of an electric train, about 1 MW, or an advertising break on television (several million electric kettles). There must therefore be a facility for varying the supply to meet sudden changes of demand. Figure 5.1 illustrates the experience of the UK Central Electricity Generating Board (CEGB). The steepness of the rise or fall in demand in a day should be noted together with the requirement to have installed capacity more than five times the minimum demand.



Sign in to download full-size image

FIG. 5.1. CEGB electrical supply.

There are two further significant characteristics of an electrical grid. First, electricity cannot effectively be stored. If, therefore, variations of demand create an imbalance between supply and demand, the difference is taken up by a change in the stored energy of the turbine-generators and any other synchronous motors, and hence in shaft speed as reflected in a change of the frequency of the supply system. Secondly, the strong electrical interconnection of the coupled motors and generators means that for our purposes all supply units act with a common inertia and common frequency change in this tightly synchronised situation.

There are many pressing reasons for keeping a constant supply frequency. Synchronous motors are used for clocks, domestically and industrially. Frequency deviation leads to a steadily increasing clock error which may be inconvenient or even dangerous. Electricity meters, on which the utility revenue depends, are thrown out of calibration by departures from standard frequencies. Other motors lose efficiency and therefore create economic problems. Overspeeding of the turbine-generator has obvious safety as well as economic implications in view of the damage that can be done and subsequent loss of availability from a blade torn loose; however, underspeeding may also be dangerous in bringing shaft speeds close to resonant frequencies below normal running speed that are usually passed through quickly in the run-up of the equipment to synchronous speed. There is, therefore, a powerful motive to keep mean fluctuations in the supply frequency down and in no case, say, to allow more than a variation of perhaps 2%. At that frequency deviation, the utility may feel obliged to disconnect the entire load with results that may approach the level of a catastrophe. Very much smaller mean square deviations are sought for which the control system must provide, of the order of 0.1%.

Not only demand fluctuates. The supply side itself is open to failures of a generating unit or an interconnection. Modern stations have a capacity of over 1000 MW(e) so that the system should be capable of adapting quickly to take up the loss of one such unit. It takes several hours to start up a reserve station from "cold" and therefore there is a necessity to have in the system a "spinning reserve" of this capacity, capable of taking up the unexpected loss as well as to react to the various demand fluctuations.

Spinning reserve to meet this requirement might therefore take the form of four 1000 MW stations operating at 75% capacity, and such stations are specified in the design stage as being able to take up an increase from, say, 75–90% load within 2 or 3 s, with the balance to 100% load in a few minutes. This specification meets the loss of supply requirement. To meet fluctuating load demands a normal station in the supply system is likely to have a specification of changing load at ± 10% a minute over at least 50% of its power range. Inevitably, then, some stations have to be run at an apparently unecomonic part load.

An ability to vary the power delivered to the turbine quickly is bound up with the existence of a reservoir of stored energy in the reactor design, particularly therefore in the heat capacity of the exchangers supplying steam in an indirect cycle or the coolant in the core or storage drum for a direct cycle. The capacity of the reactor plant to meet such a demand can be measured in the number of full-power seconds of energy stored available to meet this demand, and this in turn will be a function of the heat exchanger or boiler design. Larger capacity is purchased at a cost. There will be a trade-off point between flexibility and capital cost; the control engineer may well find himself having to argue the dynamic merits against the more static viewpoint of the reactor physicist.

We shall not go into the scheduling of power stations to meet demand changes other than to state the obvious: that stations are brought into use according to the marginal cheapness of the electricity they supply. Nuclear stations are generally high capital/low fuel cost units and therefore are used for base load. Peak demand might be met by cheaper or older, lower efficiency, fuel expensive and likely fossil-fuelled units. However, the amount of nuclear electricity now provided in many utility systems is such as to exceed the minimum base load (summer night low) and therefore of necessity, nuclear stations must be built to provide some degree of flexibility. (The UK installed nuclear capacity for the CEGB 1975/6 was 6% sent out but the proportion of nuclear electricity generated in the year was 10%.)

View chapterPurchase book

Safety guidelines for space nuclear reactor power and propulsion systems

Mohamed El-Genk, in Space Safety Regulations and Standards, 2010

Space Reactor Power Systems for Avoidance of Single-point Failures......…350

SCoRe–NaK–TE power system for global civilian air and sea traffic control satellites......…350

Startup in orbit......…352

Reactivity control and reactor shutdown......…353

Radioactivity in SCoRe core after shutdown......…354

Decay heat removal after shutdown......…355

High-Power S^4–CBC Space Reactor Power System......…355

Integration of S^4–CBC space power system......…355

Safety and reliability features......…357

Scalable AMTEC Integrated Reactor Space (SAIRS) power system......…359

View chapterPurchase book

Space Power Systems Engineering

Arthur P. Fraas, in Progress in Astronautics and Rocketry, 1966

Applications

The MPRE project is a bold attempt to obtain a mechanically simple, high-performance powerplant. If the many difficult problems involved can be solved – and we believe that we can solve them – the basic system possesses a number of inherent advantages that make it an outstanding candidate for many terrestrial applications where a highly reliable, long-life, low specific weight, compact nuclear powerplant is required. It is even conceivable that a much larger version might prove suitable for coupling to a supercritical pressure steam plant to give a binary vapor cycle that would yield an over-all thermal efficiency of about 55% for central station applications.9

View chapterPurchase book

Mainstream Power Reactor Systems

Malcolm Joyce, in Nuclear Engineering, 2018

10.3 Introduction

Nuclear power reactors use very little fuel and operate for a long time by the standards of all other base load electricity generation options: typically 30%, ~30 t of a core being replaced annually. One tangible result of this engineering trait is that while energy strategies may change, the world's stock of power reactors spans a variety of designs operating with timescales that overlap significantly. As time has passed, some designs have become the precursors of plant currently under construction, while others have been superseded but continue to make a valid and important contribution to national electricity needs, and to influence future developments. Increasing the power output of existing plant in preference to building new plant (uprating) and extending the life of existing plant (referred to as long-term operation or life extension) are important influences in this regard.

Given our focus on commercial energy production, it is important to appreciate that there is a variety of reactor systems operating throughout the world at present for research, materials irradiation and propulsion. These are based on similar engineering principles but we do not consider them further here. The first generation of power production reactors that followed the military developments after World War II (Generation I) are now shutdown and at various stages of decommissioning. A significant number of the subsequent Generation II reactor systems continue to contribute to the world's electricity needs but most are forecast to end generation by ~ 2025 with the remainder being Generation III plant built in the 1990s. The mainstream reactor designs that are options for current and future build (Generation III+) comprise the PWR, the BWR and the pressurised heavy-water reactor (PHWR) designs. The PWR and BWR are often referred to as light-water reactor (LWR) designs whereas the heritage of the PHWR design draws on the CANada Deuterium Uranium reactor design known as the CANDU. A variant of the PWR principle in use in Eastern Europe and Russia is the water–water energetic reactor (VVER). There are also reactor systems currently in operation that are unlikely to be pursued further commercially when they come to the end of their operating life, either due to economic reasons or as a result of operational experience. These comprise the GCR and the LWGR. The former is associated almost entirely with the advanced gas-cooled reactors (AGR) used in Britain while the latter with the Reaktor Bolshoy Moshchnosty Kanalny (RBMK) that are in use in Russia. A histogram of the number of operable nuclear power reactors in the world by type is provided in Fig. 10.2 with a summary of the main technical attributes provided in Table 10.1.



Sign in to download full-size image

Fig. 10.2. The number of operable thermal-spectrum fission reactors as a function of type in the world at the time of writing. PWR: pressurised water reactor; BWR: boiling water reactor; PHWR: pressurised heavy-water reactor; LWGR: light-water graphite reactor; GCR: gas-cooled reactor.

Table 10.1. A Summary of the Main Technical Attributes of the World's Current, Mainstream Operating Reactors

Design AttributePWRaBWRbGCRcPHWRdLWGReNumber of fuel channels/assemblies121–1937483323801661Number of control rods/assemblies531778925f and 28g221Core diameter/m3.44.911.07.011.8Core height/m3.74.39.86.07.0Inlet coolant temperature/°C288278339266270Outlet coolant temperature/°C326288639310284Gas flow rate/kg s− 1––4067––Operating pressure/bar160704110070Peak coolant flow rate/kg s− 117,43818201424 per ch.13,000Rated thermal power/MWth34003320162320643200Electrical power/MWe11501100600~ 675925235U enrichment/%2.1–3.10.71–3.052.2–2.70.7–2.12.0–2.6Pellet diameter/mm81014.51211.5Pellet length/mm1010–1615CladdingZircaloy-4Zircaloy-2Stainless steelZircaloy-4ZircaloyFuel element diameter/mm9.111.214.513.113.6Fuel element length/mm365837081036495.33.64No. elements per fuel assembly2645536372 × 18

a4-Loop Westinghouse design example.bGeneral Electric BWR/6 design example.cAdvanced gas-cooled reactor (AGR) example.dCanada deuterium uranium (CANDU6) example.eRBMK-1000 example.fReactor-regulating system (RRS).gShutdown system (SDS#1).

View chapterPurchase book

Introduction to the reprocessing and recycling of spent nuclear fuels

Kenneth L. Nash, Mikael Nilsson, in Reprocessing and Recycling of Spent Nuclear Fuel, 2015

1.6.2 Key process steps

When power reactor fuel is discharged from the reactor, it must first be cooled for some time (typically several years) to allow for reduction in the intensity of fission product beta and gamma radiation and coincidentally to reduce its thermal signature. After adequate thermal cooling, the fuel pins are chopped to breach the cladding material and expose the UO2 matrix. Most of the fuel dissolves in hot concentrated nitric acid, which both neutralizes the basic oxide and oxidizes susceptible components (e.g., U4 + → UO22 +). The dissolved components are separated from the undissolved residue of fuel cladding and selected fission products. In the current standard for aqueous processing, the dissolved fuel is typically contacted with one or more organic solutions to begin the separation process. Ideally, valuable components are recovered for recycle and reuse while waste is transformed to some sort of solid waste form for permanent emplacement in a geological repository.

View chapterPurchase book

Recommended publications:

Energy Research & Social Science

Journal

Browse Journals & Books



About ScienceDirect

Remote access

Shopping cart

Advertise

Contact and support

Terms and conditions

Privacy policy

Skip to Main content



Thorium Reactor

Related terms:

Thorium

Gelation

Light Water Reactors

Nuclear Energy

Molten Salt Reactor

Uranium

Reactor Design

View all Topics

Introduction

Thomas J. Dolan, in Molten Salt Reactors and Thorium Energy, 2017

1.5 Thorium fuel advantages

Approaches to a thorium reactor include:

•

Molten thorium salt fuel, such as ThF4 or ThCl4;

•

A driven subcritical molten salt system using fusion or accelerator-generated neutrons;

•

Use of a graphite moderated, He cooled pebble bed reactor;

•

The use of a seed and blanket solid fuel with an LWR cycle.

This book discusses the first two, because they involve molten salt.

The Th232–U233 breeder fuel cycle has the following advantages over the U238–Pu239 breeder fuel cycle:

•

Th232 is about four times as abundant as U238 in the earth's crust (Chapter 9: Environment, waste, and resources).

•

Almost all the thorium can be converted to fissile U-233, but it is more difficult to convert U-238 to Pu-239 with high efficiency. (See also Hargraves and Moir 2010).

•

Breeding is possible in thorium with both slow and fast neutrons. The regeneration factor η is the number of fission neutrons emitted per neutron absorbed in the fuel. Values of η≥2.2 are needed for good breeding. In 233U of η~2.24 (thermal neutrons) and 2.30 (typical fast neutron spectrum) (Fig. 1.1). In 239Pu these values are 2.01 (thermal, unsatisfactory) and 2.45 (fast), which requires a fast neutron spectrum.



Sign in to download full-size image

Figure 1.1. Variation of η with neutron energy for 233U, where η is the number of fission neutrons emitted per neutron absorbed in fuel.

Source: See also Teller, E., 1981. Editor, Fusion, Vol 1, Part B, R. W. Moir, Chapter 15, The Fusion-Fission Fuel Factory, Academic Press, NY, Teller (1981).

•

The delayed neutron fraction β is 0.0026 in 233U, compared to 0.0020 in 239Pu. The larger value of 233U makes it easier to control the 233U core.

View chapterPurchase book

Thorium molten salt reactor nuclear energy system (TMSR)

Zhimin Dai, in Molten Salt Reactors and Thorium Energy, 2017

17.2.2 Safety features of TMSR-LF

Compared to the third-generation reactors, TMSR-LF meets high level safety standards, including intrinsic safety and the ability to contain radioactivity.

TMSR-LF is intrinsically safe, which is shown in the following aspects. It operates at near atmospheric pressure, so the reactor vessel and the loops used in TMSR-LF need only low-pressure resistance. The fuel salt has higher boiling temperature (as high as 1400°C) and higher volumetric heat capacity compared to other typical coolants. Therefore, the fuel salt can absorb large amounts of heat in accidents, which will slow the progression of accidents. It is operated with negative coolant temperature reactivity coefficient, and low excessive reactivity due to the online refueling.

The safety system of TMSR-LF includes: passive residual heat removal system to cool the reactor core during an emergency, and a passive system to discharge the fuel salt to the drain tank at high temperature, which would shut down the reactor directly.

TMSR-LF has a good ability for radioactivity retention. Some accidents can be avoided, such as large break loss of coolant accident (LBLOCA), etc. Most radioactive nuclides, such as Cs-137, I-131, and Sr-90, are soluble in the salt which prevents the leaking of radioactivity to the environment. The reactor vessel and containment are used to prevent radioactivity leakage. Underground construction is adopted to prevent leaking of radioactivity, resist natural disasters, and terrorist attacks.

View chapterPurchase book

Gelation and other innovative conversion processes for aqueous-based reprocessing and recycling of spent nuclear fuels

Manuel A. Pouchon, in Reprocessing and Recycling of Spent Nuclear Fuel, 2015

14.6.3 Programs

The external gelation, especially the KFA process, is the reference production route for the TRISO particles in the HTR fuel, such as employed in the German pebble bed reactor AVR (Arbeitsgemeinschaft Versuchsreaktor Jülich) and the follow up thorium reactor THTR-300 (Thorium-Hoch-Temperatur-Reaktor) in Hamm, Germany. The process has also been used for the Japanese, American, South African, and the Chinese high temperature reactor programs. Besides these applications, the production of minor actinide-containing fuels as transmutation targets has also been addressed. One successful approach is the production of porous beads by internal gelation that can then be infiltrated with a minor actinide-containing solution. In Nästren et al. (2013), 6% americium content fuels are reported. The particles are then compacted into pellets.

View chapterPurchase book

Turning a Vision to Reality

Kelly Crandall, ... Jonathan Koehn, in Distributed Generation and its Implications for the Utility Industry, 2014

5 What's "Broken" About the Current Model

There are well-documented challenges inherent in the traditional, regulated utility model, as articulated in Chapters 11 and 14, and others that cite the recent study for the Edison Electric Institute on "disruptive challenges" to the utility industry (Kind, 2013). Energy continues to be treated as a commodity with incentives to sell more and own large, central station generation assets; utilities frequently lack transparency; they are slow-moving entities; they require a return on investment as part of the regulatory compact for providing a public good; they face difficulties with embracing renewables and new technologies because of their large infrastructure investments. Despite Xcel's remarkable advances in integrating wind energy into its portfolio, its Colorado service territory produces one of the dirtiest energy supplies in the country, with over 50% coal generation according to Xcel Energy (2013). Its baseload includes the Comanche 3 plant, which began operation in 2010 and has a life span into the 2060s.

Simultaneously, the utility industry is changing. Several chapters—Chapters 5, 9, and 16Chapter 5Chapter 9Chapter 16, in particular—note the increasing need for grid intelligence and advanced data analytics to understand and manage distributed solar, microgrids, energy storage, and electric vehicles. Futurist thinkers might point to even more dramatic changes that could be on the horizon: space-based solar power, thorium reactors, and 3D printers replacing equipment inventories. The prices of renewable resources are quickly dropping, according to Wiser and Bolinger (2013), as Xcel's recent electric resource planning process showed; the Colorado PUC approved a preferred, low-cost resource plan that included 450 MW of wind and 170 MW of utility-scale solar.7

Customers are developing widely varying expectations and providing a high quality of service and a unique customer experience will be key to managing diverse changes. Yet regulated utilities are typically required to practice a "most favored nations" policy8 by which they must treat all customers alike, regardless of local conditions that may support an alternative approach.

Unquestionably, to weather these coming changes, utilities will need to demonstrate foresight, flexibility, and responsiveness. The Boulder community began to ask: What would a utility that exhibits those traits look like? And if not Xcel, could a city-owned entity be the catalyst for this change?

View chapterPurchase book

Thorium as a nuclear fuel

Bradley S. Van Gosen, Harikrishnan Tulsidas, in Uranium for Nuclear Power, 2016

10.4.1 China

China in collaboration with the USA has very active ongoing research on thorium utilization in MSR designs (Li et al., 2015). This is a dual program involving an early solid fuel stream and advanced liquid fuel stream (WNA, 2015c). In January 2011, the China Academy of Sciences launched a research and development program on a liquid-fluoride thorium reactor (LFTR), called the thorium-breeding molten salt reactor (TMSR). In 2013, the National Energy Administration included the TMSR project among the 25 "National Energy Major Application-Technology Research and Demonstration Projects" in its "Plan of Energy Development Strategy." In 2014, the local government of Shanghai launched a major TMSR project to support the TMSR technology development.

The TMSR program is divided into three stages. In the early stage, a 10 MWt solid-fueled molten salt test reactor (TMSR- SF1) and a 2 MWt liquid-fueled molten salt experimental reactor (TMSR-LF1) are planned for construction and operation by 2016. In the engineering experimental stage, a 100 MWt solid-fueled TMSR demonstration system (TMSR-SF2) is planned by 2025 and a 10 MWt liquid-fueled molten salt experimental reactor (TMSR-LF2) is planned by 2018. The third industrial promotion stage will aim for the commercialization of a 1 GW TMSR-SF3 by 2030. A fast spectrum TMSFR-LF fast reactor optimized for burning of MA is also envisaged.

Solid fuel MSR technology was preferred in the early stage, due to the technical difficulty associated with high radioactivity of the molten salt when they contain dissolved fuels and wastes. After the accumulation of experience is gained with component design, operation, and maintenance of clean salts, use of liquid salt will be applied. Molten salt fuel is considered superior to the TRISO fuel in effectively unlimited burnup, less waste, and lower fabricating cost (WNA, 2015c).

Solid fuel envisages only partial utilization of thorium with an open fuel cycle, whereas liquid fuel designs will have a fully closed Th-U breeding cycle. Solid fuel TRISO particles will be with both low-enriched uranium and thorium, separately. The first step will be to develop solid fuel, bypassing the difficult reprocessing and refabrication options, and subsequently mastering the complex fluid fuel technology. The US cooperation with this project is primarily on the solid fuel technology, which is considered as the realistic first step.

China is developing HTR-PM, which is a graphite-moderated, helium-cooled high-temperature reactor. It is possible to use thorium in this type of reactor. Construction of a twin HTR-PM unit started in 2014, and is expected to be operational by 2017.

View chapterPurchase book

Remediation of sites contaminated by radionuclides

B.J. Merkel, M. Hoyer, in Radionuclide Behaviour in the Natural Environment, 2012

16.2.1 Uranium mining and milling sites

Main elements to be considered relevant for potential radiological exposure from uranium and thorium mining activities are uranium and thorium and elements that occur in three natural radioactive decay series (e.g. Ac, Ra, Rn Po, Bi, Tl, [Pb], Pa, At). It is clear that mining for Cu, Co, Au, Ag, Nb, rare earth elements and coal will also generate waste streams with significant amounts of radioactivity (see also Section 16.2.5) possibly exceeding permissible threshold values. Medieval silver and lead mining, for instance, resulted in numerous waste rock piles made of mining and smelting waste in the Erzgebirge, Germany. These historical heaps are archives for tracing weathering and leaching processes in mining residues including slags. Both glassy slag and barren rock from mining contain radionuclides such as 238U and 232Th. radiometric scans revealed that radioactivity exceeded the limit of 1 mSv per year in some areas (Kupsch et al., 2004).

After crushing and milling of solid ores and flotation in most cases chemical treatment is necessary. Uranium is extracted by classical techniques such as open-pit mining or deep mining. But uranium is also recovered using in situ leaching. Thorium mining is rather unimportant so far. However, thorium reactors are discussed as a promising alternative to the currently used uranium-based systems eliminating the build-up of large amounts of long-lived transuranium elements such as plutonium in the core and expanding the reach of fissile material beyond the known uranium resources. Such concepts may change mining patterns completely in the future.

If the deposit contains sulfide ores, acid mine drainage is a matter of serious concern. Sulfide is oxidised by atmospheric oxygen and Fe(III), thus forming sulfuric acid which can leach from ores containing radionuclides and toxic heavy metals. These processes result in contamination of soils and groundwater in the vicinity of the mine (Egiebor and oni, 2007). Not only mines, but also waste rock piles containing small amounts of ores can be leached by infiltrating rainwater. In many cases uranium ores are treated on site in the direct vicinity of the mine. The first step is crushing and grinding, sometimes flotation, then leaching with sulfuric acid:

UO3s+6H++3SO42−→UO2SO434−+H2O+4H+

Chemical or microbiological heap leaching is a rather simple technique producing soluble UO2(SO4)34 −. On-site and in situ leaching may lead to severe contamination of soil, the critical zone (unsaturated zone), and groundwater. The next step is purification by either solvent extraction or ion exchange. The final step is drying to ammonium diuranate, (NH4)2U2O7, which is then heated to obtain yellowcake (~ 80% U3O8 with 20% UO2 and UO3). During this process most daughter nuclides, including thorium and radium, remain in the slurry, which is deposited in tailing ponds. In many cases barium chloride is added to the tailing sludge to co-precipitate radium with barium sulfate. Therefore, huge amounts of nuclides including radium are present in tailings. Failures of tailing dams have occurred frequently in the past, leading to spilling of radioactive sludge to the downstream area (Wates, 1983). Without a proper base, sealing of the tailing pond and drainage water treatment in place, groundwater contamination is likely to occur.

Dust control is an important issue during mining operations, particularly in arid areas. This has not been done in many cases in the past, resulting in significant contamination of soil by deposited radioactive particles. Deep mining needs intensive ventilation which consequently releases huge amounts of radon to the atmosphere. Therefore, radon daughter nuclides might be found as fallout in the vicinity of ventilation shafts with 210Pb as the most prominent radionuclide.

After mine closure several options are available to administer waste rock piles, milling, treatment plants, tailings, and the mine site. Contamination of soils, critical zone, surface and groundwater has to be considered as well as waste rock piles and tailings. In situ leaching (ISL) has the advantage of extracting uranium without exposing miners to radon and radon daughter products and without creating waste rock piles and tailings. However, the disadvantage of ISL might be severe contamination of groundwater which needs clean-up.

Figure 16.1 shows the result from alpha-spectroscopy of a sample taken in a borehole at the Schneckenstein tailing (Erzgebirge, Germany). Low activities in the first 5 metres are due to the fact that the tailings were covered with heap material. Highest activities were found at a depth of 6 metres with about 9000 Bq/kg for 230Th and 226Ra. 238U content does not exceed 3000 Bq/kg but is rather high and gives evidence for the rather poor uranium extraction technology applied during the period 1947−1961. Figure 16.1 also shows that there are no significant differences in nuclide concentrations versus depth relative to each other, which would be an indication of different leaching procedures. The 238U/226Ra ratio is about 0.4 for the entire tailing; higher values were found in the cover material and at the bottom of the tailing near the contact to the natural granite basement (Merkel et al., 1998). 210Pb is slightly depleted due to some volatilisation of radon from the tailing.



Sign in to download full-size image

16.1. Specific activity of the 238U decay series nuclides versus depth of the Schneckenstein tailing at drilling location no. 2 (Merkel et al., 1998).

View chapterPurchase book

Recommended publications:

Annals of Nuclear Energy

Journal

Nuclear Engineering and Technology

Journal

Molten Salt Reactors and Thorium Energy

Book • 2017

Progress in Nuclear Energy

Journal

Browse Journals & Books



About ScienceDirect

Remote access

Shopping cart

Advertise

Contact and support

Terms and conditions

Privacy policy

We use cookies to help provide and enhance our service and tailor content and ads. By continuing you agree to the use of cookies.

Copyright © 2020 Elsevier B.V. or its licensors or contributors. ScienceDirect ® is a registered trademark of Elsevier B.V.