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Sample records for heavy water moderated reactors

  1. Heavy water moderated tubular type nuclear reactor

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  2. Method of operating heavy water moderated reactors

    Masuda, Hiroyuki.

    1980-01-01

    Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

  3. High conversion heavy water moderated reactor

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

    In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

  4. Heavy water moderated gas-cooled reactors

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  5. Heavy water moderated reactors advances and challenges

    Meneley, D.A.; Olmstead, R.A.; Yu, A.M.; Dastur, A.R.; Yu, S.K.W.

    1994-01-01

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  6. Power control device for heavy water moderated reactor

    Matsushima, Hidesuke; Masuda, Hiroyuki.

    1978-01-01

    Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

  7. Reactivity margins in heavy water moderated production reactors

    Benton, F.D.

    1981-11-01

    The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

  8. Moderator clean-up system in a heavy water reactor

    Sasada, Yasuhiro; Hamamura, Kenji.

    1983-01-01

    Purpose: To decrease the fluctuation of the poison concentration in heavy water moderator due to a heavy water clean-up system. Constitution: To a calandria tank filled with heavy water as poison-containing moderators, are connected both end of a pipeway through which heavy water flows and to which a clean-up device is provided. Strongly basic resin is filled within the clean-up device and a cooler is disposed to a pipeway at the upstream of the clean-up device. In this structure, the temperature of heavy water at the inlet of the clean-up device at a constant level between the temperature at the exit of the cooler and the lowest temperature for the moderator to thereby decrease the fluctuation in the poison concentration in the heavy water moderator due to the heavy water clean-up device. (Moriyama, K.)

  9. Fuel enrichment reduction for heavy water moderated research reactors

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  10. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    Alcala Ruiz, F.

    1984-01-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  11. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  12. Conceptual designing of reduced-moderation water reactor with heavy water coolant

    Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

    2001-12-01

    The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

  13. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Alcala Ruiz, F

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  14. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    Pinedo V, J.L.

    1979-01-01

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  15. Neutron absorption profile in a reactor moderated by different mixtures of light and heavy waters

    Nagy, Mohamed E.; Aly, Mohamed N.; Gaber, Fatma A.; Dorrah, Mahmoud E.

    2014-01-01

    Highlights: • We studied neutron absorption spectra in a mixed water moderated reactor. • Changing D 2 O% in moderator induced neutron energy spectral shift. • Most of the neutrons absorbed in control rods were epithermal. • Control rods worth changes were not proportional to changes of D 2 O% in moderator. • Control rod arrangement influenced the neutronic behavior of the reactor. - Abstract: A Monte-Carlo parametric study was carried out to investigate the neutron absorption profile in a model of LR-0 reactor when it is moderated by different mixtures of heavy/light waters at molecular ratios ranging from 0% up to 100% D 2 O at increments of 10% in D 2 O. The tallies included; neutron absorption profiles in control rods and moderator, and neutron capture profile in 238 U. The work focused on neutron absorption in control rods entailing; total mass of control rods needed to attain criticality, neutron absorption density and total neutron absorption in control rods at each of the studied mixed water moderators. The aim was to explore whether thermal neutron poisons are the most suitable poisons to be used in control rods of nuclear reactors moderated by mixed heavy/light water moderators

  16. Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2006-01-01

    As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

  17. Method of collecting helium cover gas for heavy water moderated reactor

    Miyamoto, Keiji; Ueda, Hiroshi.

    1981-01-01

    Purpose: To reduce the systematic facility cost in a heavy water moderated reactor by contriving the simplification of a helium cover gas collecting intake system. Method: A detachable low pressure metal tank and a neoprene balloon are prepared for a vacuum pump in a permanent vacuum drying facility. When all of the helium cover gas is collected from a heavy water moderated reactor, a large capacity of neoprene balloon capable of temporarily storing it under low pressure is connected to the exhaust of the vacuum pump. On the other hand, while the reactor is operating, a suitable amount of the low pressure tank or neoprene balloon is connected to the exhaust side of the pump, thereby regulating the pressure of the helium cover gas. When refeeding the cover gas, the balloon, with a large capacity for collecting and storing the cover gas is connected to the intake side of the pump. Thus, the pressure regulation, collection of all of the cover gas and refeeding of the cover gas can be conducted without using a high discharge pump and high pressure tank. (Kamimura, M.)

  18. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  19. Multidimensional space-time kinetics of a heavy water moderated nuclear reactor

    Winn, W.G.; Baumann, N.P.; Jewell, C.E.

    1980-01-01

    Diffusion theory analysis of a series of multidimensional space-time experiments is appraised in terms of the final experiment of the series. In particular, TRIMHX diffusion calculations were examined for an experiment involving free-fall insertion of a 235 U-bearing rod into a heavy water moderated reactor with a large reflector. The experimental transient flux-tilts were accurately reproduced after cross section adjustments forced agreement between static diffusion calculations and static reactor measurements. The time-dependent features were particularly well modeled, and the bulk of the small discrepancies in space-dependent features should be removable by more refined cross-section adjustments. This experiment concludes a series of space-time experiments that span a wide range of delayed neutron holdback effects. TRIMHX calculations of these experiments demonstrate the accuracy of the modeling employed in the code

  20. Improvements in gas supply systems for heavy-water moderated reactors

    Aubert, G.; Hassig, J.M.; Laurent, N.; Thomas, B.

    1964-01-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [fr

  1. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  2. Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

    Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [French] La France, qui a base son effort principal pour la production d'energie nucleaire sur la filiere des reacteurs a uranium naturel et graphite refroidis par gaz, et qui a un programme a plus

  3. Thorium in heavy water reactors

    Andersson, G.

    1984-12-01

    Advanced heavy water reactors can provide energy on a global scale beyond the foreseeable future. Their economic and safety features are promising: 1. The theoretical feasibility of the Self Sufficient Equilibrium Thorium (SSET) concept is confirmed by new calculations. Calculations show that the adjuster rod geometry used in natural uranium CANDU reactors is adequate also for SSET if the absorption in the rods is graded. 2. New fuel bundle designs can permit substantially higher power output from a CANDU reactor. The capital cost for fuel, heavy water and mechanical equipment can thereby be greatly reduced. Progress is possible with the traditional fuel material oxide, but the use of thorium metal gives much larger effects. 3. A promising long range possibility is to use pressure tanks instead of pressure tubes. Heat removal from the core is facilitated. Negative temperature and void coefficients provide inherent safety features. Refuelling under power is no longer needed if control by moderator displacement is used. Reduced quality demand on the fuel permits lower fuel costs. The neutron economy is improved by the absence of pressure and clandria tubes and also by the use of radial and axial blankets. A modular seed blanket design can reduce the Pa losses. The experience from construction of tank designs is good e.g. AAgesta, Attucha. It is now also possible to utilize technology from LWR reactors and the implementation of advanced heavy water reactors would thus be easier than HTR or LMFBR systems. (Author)

  4. Design and development of rolled joint for moderator sparger channel of an Indian Pressurised Heavy Water Reactor

    Joemon, V.; Sinha, R.K.

    1993-01-01

    Indian Pressurised Heavy Water Reactors are natural uranium fuelled heavy water moderated and cooled reactors. As per the conventional scheme, the moderator enters through one or more inlet nozzles penetrating the calandria shell and flows out through outlet nozzles. Baffles are fixed at the inlet nozzles for proper distribution of moderator in the calandria and to avoid the impact of the jet on the neighbouring calandria tubes. An alternate scheme for moderator inlet has been conceived and engineered in which three lower peripheral lattice locations of the reactor are converted into moderator inlets. This is achieved by moderator sparger channels each containing a 5 m long perforated zircaloy-2 sparger tube rolled to the calandria tube sheets and extended by stainless steel tubular components (inserts) at both ends of a sparger channel. Moderator enters the sparger channel at both ends and flows into the calandria. In the absence of standard codes for design of rolled joints, it was requires to develop these joints based on trials followed by various tests. this paper discusses the details of the rolled joint developed for this purpose, the details of the trials with test results and optimization of rolling parameters for these joints

  5. Heavy water upgrading system in the Fugen heavy water reactor

    Matsushita, T.; Susaki, S.

    1980-01-01

    The heavy water upgrading system, which is installed in the Fugen heavy water reactor (HWR) was designed to reuse degraded heavy water generated from the deuteration-dedeuteration of resin in the ion exchange column of the moderator purification system. The electrolysis method has been applied in this system on the basis of the predicted generation rate and concentration of degraded heavy water. The structural feature of the electrolytic cell is that it consists of dual cylindrical electrodes, instead of a diaphragm as in the case of conventional water electrolysis. 2 refs

  6. Technologies for tritium control in fission reactors moderated with heavy water

    Ramilo, L.B.; Gomez de Soler, S.M.

    1996-01-01

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  7. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    Chang, Se Myong; Park, A. Y.; Kim, Hyoung Tae

    2011-01-01

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  8. Neutron moderation in heavy water

    Assis, J.T. de.

    1980-03-01

    The calculation of the energetic spectrum of thermic neutrons in heavy water, according to the models of the differential cross section; is presented. Simplifications in the Butler model are suggested for the diminution of computer time. The results obtained are compared with experimental data and with the Brown - St.John model. This calculation has been done in 30 energy groups and within our limit of precision, the results with the models and simplifications present satisfactory values, allowing its inclusion in reactor codes. (Author) [pt

  9. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  10. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  11. Application of the dose limitation system to the control of carbon-14 releases from heavy-water-moderated reactors

    Beninson, D.; Gonzalez, A.J.

    1982-01-01

    Heavy-water-moderated reactors produce substantially more carbon-14 than light-water reactors. Applying the principles of the systems of dose limitation, the paper presents the rationale used for establishing the release limit for effluents containing this nuclide and for the decisions made regarding the effluent treatment in the third nuclear power station in Argentina. Production of carbon-14 in PHWR and the release routes are analysed in the light of the different effluent treatment possibilities. An optimization assessment is presented, taking into account effluent treatment and waste management costs, and the collective effective dose commitment due to the releases. The contribution of present carbon-14 releases to future individual doses is also analysed in the light of an upper bound for the contribution, representing a fraction of the individual dose limits. The paper presents the resulting requirements for the effluent treatment regarding carbon-14 and the corresponding regulatory aspects used in Argentina. (author)

  12. Heavy water cycle in the CANDU reactor

    Nanis, R.

    2000-01-01

    Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

  13. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  14. The heavy water reactors

    Brudermueller, G.

    1976-01-01

    This is a survey of the development so far of this reactor line which is in operation all over the world in various types (e.g. BHWR, PHWR). MZFR and the CANDU-type reactors are discussed in more detail. (UA) [de

  15. Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

    Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

  16. Heat exchangers in heavy water reactor systems

    Mehta, S.K.

    1988-01-01

    Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

  17. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    Chang, S.M.; Kim, H.T.

    2013-01-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  18. Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

    Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

    1977-01-01

    After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)

  19. The heavy water accountancy for research reactors in JAERI

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

    The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

  20. Advances in heavy water reactors

    1994-03-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  1. Water chemistry features of advanced heavy water reactor

    Sriram, Jayasree; Vijayan, K.; Kain, Vivekanad; Velmurugan, S.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

  2. General description of advanced heavy water reactor

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  3. Thorium utilization in heavy water moderated Accelerator Driven Systems

    Bajpai, Anil; Degweker, S.B.; Ghosh, Biplab

    2011-01-01

    Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. Presently, the volume of waste in India is small and therefore there is little incentive to develop ADS based waste transmutation technology immediately. With limited indigenous U availability and the presence of large Th deposits in the country, there is a clear incentive to develop Th related technologies. India also has vast experience in design, construction and operation of heavy water moderated reactors. Heavy water moderated reactors employing solid Th fuels can be self sustaining, but the discharge burnups are too low to be economical. A possible way to improve the performance such reactors is to use an external neutron source as is done in ADS. This paper discusses our studies on Th utilization in heavy water moderated ADSs. The study is carried out at the lattice level. The time averaged k-infinity of the Th bundle from zero burnup up to the discharge burnup is taken to be the same as the core (ensemble) averaged k-infinity. For the purpose of the analysis we have chosen standard PHWR and AHWR assemblies. Variation of the pitch and coolant (H 2 O/D 2 O) are studied. Both, the once through cycle and the recycling option are studied. In the latter case the study is carried out for various enrichments (% 233 U in Th) of the recycled Th fuel bundles. The code DTF as modified for lattice and burnup calculations (BURNTRAN) was used for carrying out the study. The once through cycle represents the most attractive ADS concept (Th burner ADS) possible for Th utilization. It avoids reprocessing of Th spent fuel and in the ideal situation the use of any fissile material either initially or for sustaining itself. The gain in this system is however rather low requiring a high power accelerator and a substantial fraction of the power generated to be fed back to the accelerator. The self sustaining Th-U cycle in a heavy moderated ADS

  4. UK methods for studying fuel management in water moderated reactors

    Fayers, F.J.

    1970-10-01

    Current UK methods for studying fuel management and for predicting the reactor physics performance for both light and heavy water moderated power reactors are reviewed. Brief descriptions are given of the less costly computer codes used for initial assessment studies, and also the more elaborate programs associated with detailed evaluation are discussed. Some of the considerations influencing the accuracy of predictions are included with examples of various types of experimental confirmation. (author)

  5. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Kulikov, G. G.; Shmelev, A. N.

    2016-01-01

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  6. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  7. A review of the UKAEA interest in heavy water reactors

    Symes, R.J.

    1983-01-01

    The chapter commences with a brief account of the history of heavy water production and then begins the story of the British use of this moderator in power reactors. This is equated with the introduction and development of the tube reactor as a distinct and important form of reactor construction in contrast with the perhaps better known vessel design that has tended to dominate reactor engineering to date. The account thus includes a succession of reactor designs including the gas and steam cooled heavy water systems in addition to the steam-generating heavy water reactor. The SGHWR was demonstrated by the construction of a substantial prototype, which continues in operation as a flexible and reliable electricity-generating plant. It was also, for a time, identified as the system to be used for Britain's third reactor programme. Today the successful Canadian CANDU power reactors represent the only penetration of heavy water reactor technology into large scale electricity generation. The range of research and experimental reactors using heavy water in their cores is reviewed. (author)

  8. Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

    Balakrishnan, M.R.

    1991-01-01

    This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

  9. Technologies for tritium control in fission reactors moderated with heavy water; Tecnologias para control de tritio en reactores de fision moderados con agua pesada

    Ramilo, L B; Gomez de Soler, S M [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors` belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the `on-line` detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs.

  10. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    Hareux, F.; Roche, R.; Vrillon, B.

    1964-01-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

  11. Method of controlling power of a heavy water reactor

    Masuda, Hiroyuki.

    1975-01-01

    Object: To adjust a level of heavy water in a region of reflection body to control power in a heavy water reactor. Structure: The interior of a core tank filled with heavy water is divided by a partition into a core heavy water region and a reflection body region formed by surrounding the core heavy water region, and a level of heavy water within the reflection body region is adjusted to control power. Preferably, it is desirable to communicate the core heavy water region with the reflection body heavy water region at their lower portion, and gas pressure applied to an upper portion within at least one of said regions is adjusted to adjust the level of heavy water within the reflection body heavy water region. Thereby, the heavy water within the reflection body heavy water region may be introduced into the core region, thus requiring no tank which stores heavy water within the reflection body region. (Kamimura, M.)

  12. Good practices in heavy water reactor operation

    2010-06-01

    The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

  13. Natural-circulation flow pattern during the gamma-heating phase of an LBLOCA in a heavy-water moderated reactor

    Rodriguez, S.B.; Unal, C.; Pasamehmetoglu, K.O.; Motley, F.E.

    1992-01-01

    In a postulated large-break loss-of-coolant accident (LBLOCA), the core of the reactor is uncovered quickly as the liquid that drains out of the tank is replaced by air. During the LBLOCA, the reactor is scrammed. the moderator tank is drained, and fuel and control rod tubes are cooled internally by forced convection via the emergency cooling system (ECS) water. However, the safety rods, reflector assemblies, tank wall, and instrument rods continue to heat up as a result of gamma deposition. These components are primarily cooled by natural/mixed convection and radiation heat transfer. In this paper, the thermal-hydraulic analysis of a reactor moderator tank exposed to air during an LBLOCA is discussed. The analysis was performed using a special version of the Transient Reactor Analysis Code (TRAC). TRAC input and code modifications considered the appropriate modeling of ECS cooling, thermal radiation heat transfer, and natural convection. The major objective of the model was to calculate the limiting component temperature (that establishes the maximum operating power) as a result of gamma heating. In addition, the nature of the moderator tank air-circulation pattern and its effects on the limiting temperature under various conditions were analyzed. None of the components were found to exceed their structural limits when the pre-scram power level was 50% of historical power

  14. 20% inlet header break analysis of Advanced Heavy Water Reactor

    Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

    2001-01-01

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  15. The thorium fuel cycle in water-moderated reactor systems

    Critoph, E.

    1977-01-01

    Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

  16. Procedure for operating a heavy water cooled power reactor

    Rau, P.; Kumpf, H.

    1981-01-01

    Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

  17. Progress in design study on reduced-moderation water reactors

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Shirakawa, Toshihisa; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takeda, Renzo [Hitachi Ltd., Tokyo (Japan); Yokoyama, Tsugio [Toshiba Corp., Kawasaki, Kanagawa (Japan); Hibi, Koki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Wada, Shigeyuki [Japan Atomic Power Co., Tokyo (Japan)

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPC) in 1998, under technical cooperation with three Japanese reactor vendors. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight-lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR type core with high void fraction and super-flat core, a long operation cycle BWR type core using void tube assembly, a high conversion BWR type core without blankets, a high conversion PWR type core using heavy water as a coolant, and a PWR type core for plutonium multi-recycle using seed-blanket type fuel assemblies. Detailed feasibility studies for the RMWR have been continued on core design study. The present report summarizes the recent progress in the design study for the RMWR. (author)

  18. Neutron disadvantage factors in heavy water and light water reactors

    Pop-Jordanov, J.

    1966-01-01

    A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

  19. Operation management of the prototype heavy water reactor 'Fugen'

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

    1983-01-01

    The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

  20. Research on Reduced-Moderation Water Reactor (RMWR)

    Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro

    1999-11-01

    The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from 238 U to 239 Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

  1. Research on Reduced-Moderation Water Reactor (RMWR)

    Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1999-11-01

    The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

  2. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  3. Operating performance of the prototype heavy water reactor Fugen

    1984-01-01

    Since the full scale operation was started in March, 1979, the ATR Fugen power station has been verifying the performance and reliability of the machinery and equipment, uranium-plutonium mixed oxide fuel and so on, and obtaining the technical prospect for putting ATRs in practical use by accumulating operation and maintenance techniques, through about five years of operation. In this report, the operational results of the Fugen power station are described. Fugen is a heavy water-moderated, boiling light water-cooled, pressure tube type reactor with 165 MWe output. As of the end of March, 1984, the total generated electric power was about 4.3 billion kWh, and the operation time was about 27,000 hours. The mean capacity ratio reached 58.8%. During the operation period, troubles including plant shutdown occurred eight times, but generally the performance and reliability of the machinery and equipment have been good. 580 fuels including 284 MOX fuels have been charged, but fuel breaking did not occur at all. The consumption of heavy water and the leak of tritium did not cause problem. The management of the core and fuel, the management of maintenance, the quality control of cooling water and heavy water, radiation control and the management of wastes are reported. (Kako, I.)

  4. Light and heavy water replacing system in reactor container

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  5. Accident analysis of heavy water cooled thorium breeder reactor

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-01-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  6. Accident analysis of heavy water cooled thorium breeder reactor

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  7. Improvements in gas supply systems for heavy-water moderated reactors; Etudes de perfectionnements aux systemes d'alimentation en gaz d'un reacteur modere a l'eau lourde

    Aubert, G; Hassig, J M; Laurent, N; Thomas, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [French] Dans un reacteur modere a l'eau lourde et refroidi au gaz sous pression, un probleme important du point de vue du trace du bloc pile et de son economie est le choix du systeme d'alimentation en gaz. Pour une solution a tubes de force, l'ensemble des structures du bloc reacteur est a temperature relativement faible, alors que les organes d'alimentation en gaz sont a celle, notablement plus elevee, du gaz. Ces organes, traverses par le debit du caloporteur, doivent lui opposer le minimum de resistance afin de ne pas necessiter un supplement onereux de puissance de

  8. Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data

    Boado, H.J.; Gho, C.J.; Abbate, M.J.

    1981-01-01

    Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author) [es

  9. Moderator for nuclear reactor

    Milgram, M.S.; Dunn, J.T.; Hart, R.S.

    1995-01-01

    This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

  10. Design of a thorium fuelled Advanced Heavy Water Reactor

    Krishnani, P.D.

    2009-01-01

    Full text: The main objective for development of Advanced Heavy Water Reactor (AHWR) is to demonstrate thorium fuel cycle technologies, along with several other advanced technologies required for next generation reactors, so that these are readily available in time for launching the third stage. The AHWR under design is a 300 MWe vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The fuel consists of (Th-Pu)O 2 and ( 233 ThU)O 2 pins. The fuel cluster is designed to generate maximum energy out of 233 U, which is bred in-situ from thorium and has a slightly negative void coefficient of reactivity, negative fuel temperature coefficient and negative power coefficient. For the AHWR, the well -proven pressure tube technology and online fuelling have been adopted. Core heat removal is by natural circulation of coolant during normal operation and shutdown conditions. Thus, it combines the advantages of light water reactors and PHWRs and removes the disadvantages of PHWRs. It has several passive safety systems for reactor normal operation, decay heat removal, emergency core cooling, confinement of radioactivity etc. The fuel cycle is based on the in-situ conversion of naturally available thorium into fissile 233 U in self sustaining mode. The uranium in the spent fuel will be reprocessed and recycled back into the reactor. The plutonium inventory will be kept a minimum and will come from fuel irradiated in Indian PHWRs. The 233 U required initially can come from the fast reactor programme or it can be produced by specially designing the initial core of AHWR using (Th,Pu)MOX fuel. There will be gradual transition from the initial core which will not contain any 233 U to an equilibrium core, which will have ( 233 U, Th) MOX fuel pins also in a composite cluster. The self sustenance is being achieved by a differential fuel loading of low and a relatively higher Pu in the composite clusters. The AHWR burns the

  11. Advanced electrolytic cascade process for tritium recovery from irradiated heavy water moderator (Preprint No. PD-15)

    Ragunathan, P.; Mitra, S.K.; Jain, D.K.; Nayar, M.G.; Ramani, M.P.S.

    1989-04-01

    The paper briefly describes a design study of an electrolytic cascade process plant for enrichment and recovery of tritium from irradiated heavy water moderators from Rajasthan Atomic Power Station Reactors. In direct multistage electrolysis process, tritiated heavy water from the reactor units is fed to the electrolytic cell modules arranged in the form of a cascade where it is enriched and decomposed into O 2 gas stream and D 2 /DT gas stream. The direct electrolysis of tritiated heavy water allows tritium to be concentrated in the aqueous phase. Several stages are used to achieve the necessary enrichment. The cascade plant incorporates the advanced electrolyser technology developed in Bhabha Atomic Research Centre (Bombay) using porous nickel electrodes, capable o f high current density operation at reduced energy consumption for electrolysis. (author). 3 tabs

  12. Possibility of using metal uranium fuel in heavy water reactors

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-01-01

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  13. The future 700 MWe pressurized heavy water reactor

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  14. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  15. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    Park, T-K.; Jung, S-H.

    1996-01-01

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  16. Nuclear calculation methods for light water moderated reactors

    Hicks, D.

    1961-02-01

    This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

  17. Reactivity requirements and safety systems for heavy water reactors

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  18. Decontamination of the RA reactor heavy water system, Annex 9

    Maksimovic, Z.B.; Nikolic, R.M.; Marinkovic, M.D.; Jelic, Lj.M.

    1963-01-01

    Both stainless steel and aluminium parts of the RA reactor heavy water system system were decontaminated as well as the heavy water itself. System was contaminated with 60 Co. Decontamination factor was determined by activity measurements during distillation. Concentration of the corrosion products in the heavy water was measured by spectrochemical analysis, and found to be 0.1 - 1 mg/l. Chemical analyses of the aluminium and stainless steel surfaces showed that cobalt was adsorbed on the aluminium oxide layer. Water solution of 7%H 3 PO 4 + 2% CrO 3 was used for decontamination of the heavy water system and distillation device. This was found to be the most efficient solvent which does not affect stainless steel corrosion. Decontamination factors achieved were from 60 - 100. Decontamination results enabled determining the distribution of cobalt in the system: 10 Ci on the stainless steel parts, 50 Ci in the heavy water; and above 600 Ci on the fuel and experimental channels. Specific activity of 60 Co was calculated to be 15 Ci/g on the reactor channels, 8 Ci/g on the stainless steel parts and 3 Ci/g in the heavy water. Decontamination of the aluminium parts was not done because it was considered it could initiate corrosion. Since the efficiency of distillation is increased it was expected that permanent distillation would remove most of the activity in the reactor channels

  19. Conceptual design of a pressure tube light water reactor with variable moderator control

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-01-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  20. Study on core design for reduced-moderation water reactors

    Okubo, Tsutomu

    2002-01-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  1. Study on core design for reduced-moderation water reactors

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  2. FLUID MODERATED REACTOR

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  3. Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

    Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

    2013-01-01

    The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

  4. The steam generating heavy water reactor

    Middleton, J.E.

    1975-01-01

    A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

  5. Possibilities of using metal uranium fuel in heavy water reactors

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-11-01

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

  6. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  7. The thorium fuel cycle in water-moderated reactor systems

    Critoph, E.

    1977-05-01

    Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An 'inventory' of uranium of between 1 and 2 Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium) is some two decades

  8. Experience in operation of heavy water reactors

    Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

    1999-01-01

    The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

  9. Conceptual design of a large heavy water reactor for US siting

    Shapiro, N.L.; Jesick, J.F.

    1979-09-01

    Information is presented concerning fuel management and safety and licensing assessment of the pressurized heavy water reactor; and commercial introduction of the pressurized heavy water reactor in the United States

  10. Transmutation of Americium in Light and Heavy Water Reactors

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

    2009-06-15

    There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

  11. The Steam Generating Heavy Water Reactor

    Middleton, J.E.

    1975-01-01

    An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

  12. Heavy-Water Power Reactors. Proceedings Of A Symposium

    NONE

    1968-04-15

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  13. Heavy-Water Power Reactors. Proceedings Of A Symposium

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  14. Pressurized heavy-water reactor safety

    Pease, L.; Wilson, R.

    1977-09-01

    CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

  15. Comparison of methods for the determination of boron in heavy water moderator

    Green, L.W.; Davey, E.C.; Gulens, J.; Longhurst, T.H.; Mislan, J.P. (Atomic Energy of Canada Ltd., Chalk River, Ontario. Chalk River Nuclear Labs.)

    1984-08-01

    Five analysis methods were compared for the determination of boron in heavy water moderator: isotope dilution mass spectrometry, spectrophotometry, neutron activation, inductively coupled plasma -atomic emission spectrometry, and ion selective electrode potentiometry. Ten samples were analysed by each method; the results showed close agreement between all of the methods. Only mass spectrometry achieved the required precision (<1 percent rsd) for samples taken during initial reactor operation, but all of the methods achieved sufficient precision (<10 percent rsd) for samples taken during normal operation. For samples for which the /sup 10/B concentration must be determined, only mass spectrometry and neutron activation are applicable.

  16. Heavy water and nonproliferation

    Miller, M.M.

    1980-05-01

    This report begins with a historical sketch of heavy water. The report next assesses the nonproliferation implications of the use of heavy water-moderated power reactors; several different reactor types are discussed, but the focus is on the natural uranium, on-power fueled, pressure tube reactor CANDU. The need for and development of on-power fueling safeguards is discussed. Also considered is the use of heavy water in plutonium production reactors as well as the broader issue of the relative nuclear leverage that suppliers can bring to bear on countries with natural uranium-fueled reactors as compared to those using enriched designs. The final chapter reviews heavy water production methods and analyzes the difficulties involved in implementing these on both a large and a small scale. It concludes with an overview of proprietary and nonproliferation constraints on heavy water technology transfer

  17. Heavy water reactors on the once-through uranium cycle

    1978-05-01

    This paper presents preliminary technical and economic data to INFCE on the once-through uranium fuel cycle for use in early comparisons of alternate nuclear systems. The denatured thorium fuel cycle is discussed in a companion paper. Information for this paper was developed under an ongoing program, and more complete reporting of the evaluation of the heavy water reactor and its fuel cycles is planned toward the end of the year

  18. Topical papers on heavy water, fuel fabrication and reactors

    1978-01-01

    A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

  19. Core construction in a pressure tube type heavy water reactor

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  20. Advances in commercial heavy water reactor power stations

    Brooks, G.L.

    1987-01-01

    Generating stations employing heavy water reactors have now firmly established an enviable record for reliable, economic electricity generation. Their designers recognize, however, that further improvements are both possible and necessary to ensure that this reactor type remains attractively competitive with alternative nuclear power systems and with fossil-fuelled generation plants. This paper outlines planned development thrusts in a number of important areas, viz., capital cost reduction, advanced fuel cycles, safety, capacity factor, life extension, load following, operator aida, and personnel radiation exposure. (author)

  1. New fuel advanced heavy water reactors

    Notari, Carla

    1999-01-01

    A redesign of the PHWR fuel element (FE) to be used in all Argentine nuclear power plants has been proposed elsewhere. This new FE presents several characteristics aimed to an improved in-core performance and economical benefits derived from the unification of most of the fabrication processes that today constitute two different production lines: one for Embalse nuclear power plant CANDU type fuel and another for Atucha I. Atucha I and Embalse, the two operating nuclear power plants in Argentina, are PHWR of different conception. Atucha I (357 M we) is of pressure vessel type and the fuel elements are full-length assemblies (530 cm of active length) with 36 uranium rods in the cluster and a support one in the outer ring. Embalse (648 M we) is a CANDU pressure tube reactor fuelled with the well known 37 rod / 50 cm length fuel bundles, twelve of which are loaded in each channel. The more relevant changes in the proposed design are an increased subdivision of the fuel material in 52 rods and a 100 cm long bundle. The combined features give the adequate channel pressure drop. The proposed CARA design shows a superior neutronic performance than the standard PHWR fuel elements currently used in Atucha I and Embalse nuclear power plants. A variant of the CARA FE consisting in the elimination of the central four rods, leaving 48 rods and a central free space, is strongly recommended because it saves materials (less uranium, less sheaths) with no loss of burnup. The central D 2 O zone allows a better utilization of the inner rods and compensates the diminished uranium loading. In Embalse no differences in core physics are expected except the beneficial decrease in linear power density. In Atucha I besides the lower power density, a higher exit burnup appears as a consequence of the higher uranium inventory. The exit burnup figures have been calculated with cell and reactor models and the result is that similar fuel management schemes as the proposed for Atucha I for the

  2. Chemical elimination of alumina in suspension in nuclear reactors heavy water

    Ledoux, A.

    1967-02-01

    Corrosion of aluminium in contact with moderating water in nuclear reactor leads to the formation of an alumina hydrosol which can have an adverse effect on the operation of the reactor. Several physical methods have been used in an attempt to counteract this effect. The method proposed here consists in the elimination of the aluminium by dissolution and subsequent fixation in the ionic form on mixed-bed ion-exchange resin. In order to do this, the parameters and the values of these parameters most favorable to the dissolution process have been determined. If the moderator is heavy water, the deuterated acid can be prepared by converting a solution in heavy water to a salt of the acid using a deuterated cationic resin. (author) [fr

  3. Core design study on reduced-moderation water reactors

    Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

    2002-01-01

    The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

  4. Liquid wall boiler and moderator (BAM) for heavy ion-pellet fusion reactors

    Powell, J.R.; Lazareth, O.; Fillo, J.

    1977-11-01

    Thick liquid wall blankets appear to be of great promise for heavy ion pellet fusion reactors. They avoid the severe problems of intense radiation and blast damage that would be encountered with solid blanket structures. The liquid wall material can be chosen so that its vapor pressure at the working temperature of the power cycle is well below the value at which it might interfere with the propagation of the heavy ion beam. The liquid wall can be arranged so that it does not contact any surrounding solid structure when the pellet explosion occurs, including the ends. The ends can be magnetically closed just before the pellet explosion, or a time phased flow can be used, which will leave a clear central zone into which the pellet is injected. Parametric analysis comparing three candidate liquid wall materials were carried out. The three materials were lithium, flibe, and lead (with a low concentration of disolved lithium). Lead appeared to be the best choice for the liquid wall, although any of the three should allow a practical reactor system. The parametric analyses examined the effects of pellet yield (0 to 10 GJ), pellet mass (3 g to 3 kg), liquid wall thickness (10 cm to 80 cm), vapor condensation time (0 to 10 milliseconds), degree of neutron moderation in the pellet (none to 100%), liquid wall chamber size (radius of 1.5 meters to 4 meters), Pb/Li 6 ratio (100 to 5,000), and thickness of graphite moderating zone behind the liquid wall

  5. Role of passive valves & devices in poison injection system of advanced heavy water reactor

    Sapra, M.K.; Kundu, S.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

    2014-01-01

    The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water (H 2 O) cooled, heavy water (D 2 O) moderated reactor. The reactor design is based on well-proven water reactor technologies and incorporates a number of passive safety features such as natural circulation core cooling; direct in-bundle injection of light water coolant during a Loss of Coolant Accident (LOCA) from Advanced Accumulators and Gravity Driven Water Pool by passive means; Passive Decay Heat Removal using Isolation Condensers, Passive Containment Cooling System and Passive Containment Isolation System. In addition to above, there is another passive safety system named as Passive Poison Injection System (PPIS) which is capable of shutting down the reactor for a prolonged time. It is an additional safety system in AHWR to fulfill the shutdown function in the event of failure of wired shutdown systems i.e. primary and secondary shut down systems of the reactor. When demanded, PPIS injects the liquid poison into the moderator by passive means using passive valves and devices. On increase of main heat transport (MHT) system pressure beyond a predetermined value, a set of rupture disks burst, which in-turn actuate the passive valve. The opening of passive valve initiates inrush of high pressure helium gas into poison tanks to push the poison into the moderator system, thereby shutting down the reactor. This paper primarily deals with design and development of Passive Poison Injection System (PPIS) and its passive valves & devices. Recently, a prototype DN 65 size Poison Injection Passive Valve (PIPV) has been developed for AHWR usage and tested rigorously under simulated conditions. The paper will highlight the role of passive valves & devices in PPIS of AHWR. The design concept and test results of passive valves along with rupture disk performance will also be covered. (author)

  6. Growth scenarios with thorium fuel cycles in pressurised heavy water reactors

    Balakrishnan, M.R.

    1991-01-01

    Since India has generous deposits of thorium, the availability of thorium will not be a limiting factor in any growth scenario. It is fairly well accepted that the best system for utilisation of thorium is the heavy water reactor. The growth scenarios possible using thorium in HWRs are considered. The base has been taken as 50,000 tons of natural uranium and practically unlimited thorium. The reference reactor has been assumed to be the PHWR, and all other growth scenarios are compared with the growth scenario provided by the once-through natural cycle in the PHWR. Two reactor types have been considered: the heavy water moderated, heavy water cooled, pressure tube reactor, known as the PHWR; and the heavy water moderated and cooled pressure vessel kind, similar to the ATUCHA reactor in Argentina. For each reactor, a number of different fuel cycles have been studied. All these cycles have been based on thorium. These are: the self-sustaining equilibrium thorium cycle (SSET); the high conversion ratio high burnup cycle; and the once through thorium cycle (OTT). The cycle have been initiated in two ways: one is by starting the cycle with natural uranium, reprocessing the spent fuel to obtain plutonium, and use that plutonium to initiate the thorium cycle; the other is to enrich the uranium to about 2-3% U-235 (the so-called Low Enriched Uranium or LEU), and use the LEU to initiate the thorium cycle. Both cases have been studied, and growth scenarios have been projected for every one of the possible combinations. (author). 1 tab

  7. Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

    Norinder, O

    1960-11-15

    The set of neutron kinetic equations developed in this report permits the use of long integration steps during stepwise integration. Thermal relations which describe the transfer of heat from fuel to coolant are derived. The influence upon the kinetic behavior of the reactor of a number of parameters is studied. A comparison of the kinetic properties of the hot and cold moderators is given.

  8. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  9. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  10. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  11. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    Brand, P.; Wiechers, W.K.

    1974-01-01

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  12. Natural uranium fueled light water moderated breeding hybrid power reactors

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  13. High converter pressurized water reactor with heavy water as a coolant

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  14. Future development in heavy water reactors in Canada

    Donnelly, J.; Hart, R.G.

    1982-01-01

    1982 marks the 35th anniversary of the start-up of Canada's first research and test reactor, NRX. Its first power reactor has been operating successfully for the past 20 years. With 5,000 MWe of domestic capacity installed, Canada's major CANDU (Canada Deuterium, Uranium) nuclear program has a further 9,500 MWe under construction in Canada for completion by 1990 as well as committed offshore projects in Argentina, Korea and Romania. The CANDU operating record, by any measure of performance, has been outstanding. This performance is largely due to the discipline imposed on the development, design, construction and operation by two fundamental choices: natural uranium and heavy water. The impact of these two choices on availability, fuel utilization, safety and economics is discussed. Future plans call for building on those characteristics which have made CANDU so successful. When time for change comes, current assessments indicate that it will be possible to convert to more efficient advanced fuel cycles without major changes to the basic CANDU design. Primary attention is being focussed on thorium fuel cycles to ensure an abundant and continuing supply of low cost energy for the long term. The resource savings available from these fuel cycles in expanding systems are reviewed and compared with those available from LWR's and Fast Breeders. The results clearly illustrate the versatility of the CANDU reactor. It can benefit from enrichment plants or get along without them. It can complement LWR's or compete with them. It can complement Fast Breeder Reactors or compete with them as well. In the very long term CANDU's and Fast Breeders combined offer the potential of burning all the world's uranium and all the world's thorium. (author)

  15. Breeding capability and void reactivity analysis of heavy-water-cooled thorium reactor

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2008-01-01

    The fuel breeding and void reactivity coefficient of thorium reactors have been investigated using heavy water as coolant for several parametric surveys on moderator-to-fuel ratio (MFR) and burnup. The equilibrium fuel cycle burnup calculation has been performed, which is coupled with the cell calculation for this evaluation. The η of 233 U shows its superiority over other fissile nuclides in the surveyed MFR ranges and always stays higher than 2.1, which indicates that the reactor has a breeding condition for a wide range of MFR. A breeding condition with a burnup comparable to that of a standard PWR or higher can be achieved by adopting a larger pin gap (1-6 mm), and a pin gap of about 2 mm can be used to achieve a breeding ratio (BR) of 1.1. A feasible design region of the reactors, which fulfills the breeding condition and negative void reactivity coefficient, has been found. A heavy-water-cooled PWR-type Th- 233 U fuel reactor can be designed as a breeder reactor with negative void coefficient. (author)

  16. Benchmarking severe accident computer codes for heavy water reactor applications

    Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

    2010-07-01

    Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

  17. The key design features of the Indian advanced heavy water reactor

    Sinha, R.K.; Kakodkar, A.; Anand, A.K.; Venkat Raj, V.; Balakrishnan, K.

    1999-01-01

    The 235 MWe Indian Advanced Heavy Water Reactor (AHWR) is a vertical, pressure tube type, boiling light water cooled reactor. The three key specific features of design of the AHWR, having a large impact on its viability, safety and economics, relate to its reactor physics, coolant channel, and passive safety features. The reactor physics design is tuned for maximising use of thorium based fuel, and achieving a slightly negative void coefficient of reactivity. The fulfilment of these requirements has been possible through use of PuO 2 -ThO 2 MOX, and ThO 2 -U 233 O 2 MOX in different pins of the same fuel cluster, and use of a heterogeneous moderator consisting of pyrolytic carbon and heavy water in 80%-20% volume ratio. The coolant channels of AHWR are designed for easy replaceability of pressure tubes, during normal maintenance shutdowns. The removal of pressure tube along with bottom end-fitting, using rolled joint detachment technology, can be done in AHWR coolant channels without disturbing the top end-fitting, tail pipe and feeder connections, and all other appendages of the coolant channel. The AHWR incorporates several passive safety features. These include core heat removal through natural circulation, direct injection of Emergency Core Coolant System (ECCS) water in fuel, passive systems for containment cooling and isolation, and availability of a large inventory of borated water in overhead Gravity Driven Water Pool (GDWP) to facilitate sustenance of core decay heat removal, ECCS injection, and containment cooling for three days without invoking any active systems or operator action. Incorporation of these features has been done together with considerable design simplifications, and elimination of several reactor grade equipment. A rigorous evaluation of feasibility of AHWR design concept has been completed. The economy enhancing aspects of its key design features are expected to compensate for relative complexity of the thorium fuel cycle activities

  18. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  19. A case study for INPRO methodology based on Indian advanced heavy water reactor

    Anantharaman, K.; Saha, D.; Sinha, R.K.

    2004-01-01

    Under Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a methodology (INPRO methodology) has been developed which can be used to evaluate a given energy system or a component of such a system on a national and/or global basis. The INPRO study can be used for assessing the potential of the innovative reactor in terms of economics, sustainability and environment, safety, waste management, proliferation resistance and cross cutting issues. India, a participant in INPRO program, is engaged in a case study applying INPRO methodology based on Advanced Heavy Water Reactor (AHWR). AHWR is a 300 MWe, boiling light water cooled, heavy water moderated and vertical pressure tube type reactor. Thorium utilization is very essential for Indian nuclear power program considering the indigenous resource availability. The AHWR is designed to produce most of its power from thorium, aided by a small input of plutonium-based fuel. The features of AHWR are described in the paper. The case study covers the fuel cycle, to be followed in the near future, for AHWR. The paper deals with initial observations of the case study with regard to fuel cycle issues. (authors)

  20. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  1. Removal of aluminum turbidity from heavy water reactors by precipitation ion exchange using magnesium hydroxide

    Venkateswarlu, K.S.; Shanker, R.; Velmurugan, S.; Venkateswaran, G.; Rao, M.R.

    1988-01-01

    A special magnesium hydroxide MG(OH)/sub 2/ sorber, loaded onto an ion-exchange matrix has been developed to remove hydrated alumina turbidity in heavy water. This sorber was applied to the coolant/moderator system in the research reactor Dhruva. The sorber not only removed turbidity but also suspended uranium at parts per billion levels and associated β, γ activity. The sorption is based on the attraction between the positively charged Mg(OH)/sub 2/ surface and the negatively charged hydrated alumina particles

  2. Plant life management processes and practices for heavy water reactors

    Kang, K.-S.; Cleveland, J.; Clark, C.R.

    2006-01-01

    In general, heavy water reactor (HWR) nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions are depending on essentially business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. Continued plant operation, including operation beyond design life, called 'long term operation, depends, among other things, on the material condition of the plant. This is influenced significantly by the effectiveness of ageing management. Key attributes of an effective plant life management program include a focus on important systems, structure and components (SSCs) which are susceptible to ageing degradation, a balance of proactive and reactive ageing management programmes, and a team approach that ensures the co-ordination of and communication between all relevant nuclear power plant and external programmes. Most HWR NPP owners/operators use a mix of maintenance, surveillance and inspection (MSI) programs as the primary means of managing ageing. Often these programs are experienced-based and/or time-based and may not be optimised for detecting and/or managing ageing effects. From time-to-time, operational history has shown that this practice can be too reactive, as it leads to dealing with ageing effects (degradation of SSCs) after they have been detected. In many cases premature and/or undetected ageing cannot be traced back to one specific reason or an explicit error. The root cause is often a lack of communication, documentation and/or co-ordination between design, commissioning, operation or maintenance organizations. This lack of effective communication and interfacing frequently arises because, with the exception of major SSCs, such as the fuel channels or steam generators, there is a lack of explicit

  3. Heavy water reactors physics; Physique des reacteurs a eau lourde

    Girard, Y; Lourme, P; Naudet, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors) [French] Un important programme d'etudes sur la physique des reacteurs a eau lourde est mene en France depuis assez longtemps. La decision de construire le prototype EL 4 et de s'engager ainsi dans la filiere des reacteurs a eau lourde refroidis par gaz a redonne un nouvel interet a ce programme et l'a en meme temps oriente dans une direction plus particuliere. La presente communication, rassemble les resultats des etudes faites dans ce domaine depuis la derniere conference de Geneve. Dans la premiere partie on decrit les etudes experimentales dont la plupart ont ete effectuees dans la pile d'experiences critiques Aquilon II. Les experiences sont groupees en quatre ensembles: etude systematique de reseaux (mesures de laplaciens) etudes

  4. Power spectral density measurements with 252Cf for a light water moderated research reactor

    King, W.T.; Mihalczo, J.T.

    1979-01-01

    A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

  5. Heavy water upgrader of 'Fugen'

    Matsushita, Tadashi; Sasaki, Shigeo

    1980-01-01

    The nuclear power station of the advanced thermal prototype reactor ''Fugen'' has continued the smooth operation since it started the fullscale operation in March, 1979. Fugen is the first heavy water-moderated, boiling light water-cooled reactor in Japan, and its outstanding feature is the use of heavy water as the moderator. The quantity of heavy water retained in Fugen is about 140 m 3 , and the concentration is 99.8 wt.%. This heavy water had been made is USA, and was imported from F.R. of Germany where it had been used. Heavy water is an internationally regulated material, and it is very expensive and hard to purchase. Therefore in order to prevent the deterioration of heavy water and to avoid its loss as far as possible, the management of the quantity and the control of the water quality have been carried out carefully and strictly. The generation of deteriorated heavy water occurs from the exchange of ion exchange resin for poison removal and purification. The heavy water upgrader reconcentrates the deteriorated heavy water of high concentration and returns to the heavy water system, and it was installed for the purpose of reducing the purchase of supplementary heavy water. The outline of the heavy water upgrader, its construction, the performance test and the operation are described. (Kako, I.)

  6. Operation management of the prototype heavy water reactor 'Fugen'

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

    1983-09-01

    The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported.

  7. Thorium fuels for heavy water reactors. Romanian experience

    Glodeanu, F.; Mirion, I.; Mehedinteanu, S.; Balan, V.

    1984-01-01

    The renewed interest in thorium fuel cycle due to the increased demand for fissile materials has resulted in speeding up the related research and development activities. For heavy water reactors the thorium cycles, especially SSET, are very promising and many efforts are made to demonstrate their feasibility. In our country, at INPR, the research and development activity has been initiated in the following areas: the conceptual design of thorium bearing fuel elements; fuel modelling; nuclear grade thorium dioxide powder technology; mixed oxide fuel technology. In the design area, the key factors in performance limitation, especially at extended burnup have been accounted and different remedies proposed. An irradiation programme has been settled and will start this year. The modelling activities are focused on mixed oxide behaviour and material data measurements are in progress. In the nuclear grade thorium powder technology area, a good piece of work has been done to develop an integrated technology for monasite processing (thorium being a by-product in lanthanides extraction). As regards the mixed oxide fuel technology, efforts have been made to obtain (ThU)O 2 pellets with good homogeneity and high density at different compositions. Besides the mixing powders route, other non-conventional technologies for refabrication like: microspheres, pellet impregnation and clay extrusion are studied. Experimental fuel rods for irradiation testing have been manufactured. (author)

  8. Evaluation of A-1 reactor heavy-water calandria specimens

    Brumovsky, M.

    1976-01-01

    Container chains with surveillance specimens were placed in two special channels of the core peripheral part to test changes in mechanical properties due to reactor operation of caisson tube material. The specimens were made from the caisson tube material and placed by eight pieces on the outer surface of the containers. The first removed specimens were tested for corrosion losses, tensile strength, and fractured surfaces were then assessed. The changes in strength properties were found to be similar in both base material and welded joints. The corrosion film on surveillance specimens did not practically affect strength properties nor ductility. It was found that the Al-Mg-Si alloy used for the heavy water vessel caisson tubes following stabilization annealing was fully stable at operating temperatures of up to 100 degC. Slio.ht changes in properties can be attributed to the effect of a high neutron dose. Thus, the high radiation and temperature stability of the alloy was confirmed. (O.K.)

  9. Start-up test of the prototype heavy water reactor 'FUGEN', (1)

    Ando, Hideki; Kawahara, Toshio

    1982-01-01

    The advanced thermal prototype reactor ''Fugen'' is a heavy water-moderated, boiling light water-cooled power reactor with electric output of 165 MW, which has been developed since 1966 as a national project. The start-up test was begun in March, 1978, being scheduled for about one year, and in March, 1979, it passed the final pre-use inspection and began the full scale operation. In this paper, the result of the start-up test of Fugen is reported. From the experience of the start-up test of Fugen, the following matters are important for the execution of start-up test. 1) Exact testing plan and work schedule, 2) the organization to perform the test, 3) the rapid evaluation of test results and the reflection to next testing plan, and 4) the reflection of test results to rated operation, regular inspection and so on. In the testing plan, the core characteristics peculiar to Fugen, and the features of heavy water-helium system, control system and other equipment were added to the contents of the start-up test of BWRs. The items of the start-up test were reactor physics test, plant equipment performance test, plant dynamic characteristic test, chemical and radiation measurement, and combined test. The organization to perform the start-up test, and the progress and the results of the test are reported. (Kako, I.)

  10. Investigation of the heavy water distillation system at the RA reactor

    Zecevic, V.; Badrljica, R.

    1963-01-01

    The heavy water distillation system was tested because this was not done before the reactor start-up. Detailed inspection of the system components showed satisfactory results. Leak testing was done as well as the testing of the instrumentation which enables reliable performance of the system. Performance testing was done with ordinary water and later 2700 l of heavy water from the reactor was purified, decreasing the activity by 45%

  11. Construction management of Indian pressurized heavy water reactors

    Bohra, S.A.; Sharma, P.D.

    2006-01-01

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  12. Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

    De, S.K.; Karhadkar, C.G.

    2017-01-01

    An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

  13. Current developments and future challenges in physics analyses of the NRU heavy water research reactor

    Nguyen, S.; Wilkin, B.; Leung, T., E-mail: nguyens@aecl.ca, E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    The National Research Universal (NRU) reactor is heavy water cooled and moderated, with on-power fueling capability. TRIAD, a 3D two-group diffusion code, is currently used for support of day-to-day NRU operations. Recently, an MCNP full reactor model of NRU has been developed for benchmarking TRIAD. While reactivity changes and flux and power distributions from both methods are in reasonably good agreement, MCNP appears to eliminate a k-eff bias in TRIAD. Beyond TRIAD's capability, MCNP enables the assessment of radiation in the NRU outer structure. Challenges include improving TRIAD accuracy and MCNP performance, as well as performing NRU core-following using MCNP. (author)

  14. Probablistic risk assessment methodology application to Indian pressurised heavy water reactors

    Babar, A.K.; Grover, R.B.; Mehra, V.K.; Gangwal, D.K.; Chakraborty, G.

    1987-01-01

    Probabilistic risk assessment in the context of nuclear power plants is associated with models that predict the offsite radiological releases resulting from reactor accidents. Level 1 PRA deals with the identification of accident sequences relevant to the design of a system and also with their quantitative estimation. It is characterised by event tree, fault tree analysis. The initiating events applicable to pressurised heavy water reactors have been considered and the dominating initiating events essential for detailed studies are identified in this paper. Reliability analysis and the associated problems encountered during the case studies are mentioned briefly. It is imperative to validate the failure data used for analysis. Bayesian technique has been employed for the same and a brief account is included herein. A few important observations, e.g. effects of the presence of moderator, made during the application of probabilistic risk assessment methodology are also discussed. (author)

  15. Heavy water isotopic rectification in the ''ORPHEE'' reactor. SACLAY studies Centre

    Lejeune, P.; Breant, P.

    1993-01-01

    ORPHEE reactor supplies neutron beams, which are got back in a heavy water reflector. The neutron beams intensity depends on the reflector quality which is determined by the isotopic content of the heavy water. The deuterium submitted to core irradiation changes in radioactive tritium which must be eliminated largely for reasons of safety. The column must keep the heavy water isotopic content of the reflector to a value higher than 99.8% by eliminating light water by fractional distillation or rectification. This column is also used for the tritium elimination of heavy water. 13 figs

  16. Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11

    Nikolic, M.

    1964-01-01

    The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was ±1 cm

  17. Summary of the 4th workshop on the reduced-moderation water reactor

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  18. Summary of the 4th workshop on the reduced-moderation water reactor

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  19. Summary of the 3rd workshop on the reduced-moderation water reactor

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  20. Summary of the 3rd workshop on the reduced-moderation water reactor

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  1. Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

    Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

    1968-04-15

    Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

  2. Development of the heavy-water organic-cooled reactor. Status report from the United States of America

    Trilling, C A [Atomics International, Division of North American Aviation, Inc., Canoga Park, CA (United States)

    1967-01-01

    In late 1964 the United States Atomic Energy Commission decided to undertake the development of the heavy-water-moderated nuclear power reactor as part of its overall programme for the development of advanced converter reactors. The inclusion of the heavy-water reactor concept was based on its indicated potential for achieving: efficient utilization of available fuel resources; generation of low cost electric power; feasibility of scale-up to very large single unit plant sizes for the dual purpose of generating power and desalting sea water. The excellent neutron economy inherent in heavy-water moderation allows a significant increase in the amount of power which can be generated from a given amount of ore. If one takes into account the amount of uranium required not only for burn-up but also to inventory new reactors in a rapidly expanding nuclear economy, heavy-water reactors show the potential of extracting one and a half to two times more power from the ore mined than light-water reactors. Such an improvement in dynamic fuel utilization will postpone the depletion of low cost uranium ore reserves, providing more time for the discovery of new ore resources and the development of economic fast breeder reactors. The excellent neutron economy of the heavy-water reactor also allows the achievement of appreciable burn-up with low enrichment fuel, with consequent low fuel cycle costs and therefore low energy generation costs. These low fuel cycle costs make the economics of this type of reactor rather insensitive to rising ore costs. They also make the concept well suited for the most economic production of the large quantities of heat required for water desalination. The use of individual pressure tubes for circulating the coolant through the reactor vessel lends itself to the development of a modular type design, which can be scaled up to very large single unit plant sizes by simply increasing the number of identical pressure tube modules and the number of coolant

  3. Determination of reactor parameters in a simulated RA reactor lattice by measuring the reactivity level of heavy water in D{sub 2}O moderated RB reactor; Odredjivanje reaktorskih parametara u simuliranoj resetki reaktora RA merenjem reaktivnosti nivoa teske vode u D{sub 2}O moderiranom reaktoru RB

    Takac, S M; Markovic, H D; Dimitrijevic, Z B [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1974-07-01

    Direct method for experimental determination of the neutron age {tau} in the reactor lattice is not developed. Fermi theory for determining {tau} by measuring the reactivity level of the heavy water can be applied for a limited number of reactor lattices. An attempt was made to apply this approach for a complex reactor core with side and upper reflector. As expected the obtained results were not satisfactory, and {tau} was determined by Dessauer formula which gives more realistic estimation of thermalization in the reactor cell. major discrepancies are resulting from the fact that the lower reflector was neglected. But it is possible to to determine {tau} for reactor core with reflectors by additional measurements of axial distribution and other experimental data for {tau}. This is quite tedious numerical procedure. Obtained experimental data for a number of reactor parameters are compared to the initial data of the RA reactor core showing very good agreement. [Serbo-Croat] Ekspeimentalna metoda za direktno odredjivanje starosti neutrona {tau}, u resetki reaktora do danas nije razvijena. Primenjena Fermijeova teorija za odredjivanje {tau} preko merenja reaktivnosti nivoa teske vode, ocigledno se moze koristiti samo na ogranicen spektar reaktorskih resetki. U ovom radu pokusano je da se vidi primenljivost iste u slozenom - bocno i odozdo - reflektovanom jezgru reaktora. Naravno dobijeni rezultati nisu zadovoljili, sto se i moglo ocekivati, pa je {tau} odredjen preko Dessauer-ove formule, sto daje daleko blizu sliku stvarnog stanja procesa termalizacije u celiji reaktora. Ocigledno vece neslaganje primenjene teorije dolazi od zanemarivanja donjeg reflektora u jezgru reaktora. Medjutim, dopnskim merenjem aksijalne raspodele i na osnovu ostalih eksperimentalnih podataka za {tau}, moguce je odrediti eksperimentalno reflektorski koeficijent za visestruko reflektovano jezgro reaktora, sto numericki predstavlja mukotrpan i dugotrajan rad na racunaru. Dobijeni

  4. Noise analysis method for monitoring the moderator temperature coefficient of pressurized water reactors: Neural network calibration

    Thomas, J.R. Jr.; Adams, J.T.

    1994-01-01

    A neural network was trained with data for the frequency response function between in-core neutron noise and core-exit thermocouple noise in a pressurized water reactor, with the moderator temperature coefficient (MTC) as target. The trained network was subsequently used to predict the MTC at other points in the same fuel cycle. Results support use of the method for operating pressurized water reactors provided noise data can be accumulated for several fuel cycles to provide a training base

  5. Loss of coolant analysis for CIRENE-LATINA heavy water reactor

    Chiantore, B.; Dubbini, M.; Proto, G.

    1978-01-01

    CIRENE is a heavy-water moderated, boiling water cooled pressure tube reactor. Fuel is natural uranium. A variety of breaks in the primary coolant system have been postulated for the analysis of the CIRENE Latina Plant (now under construction) such as double-end break of inlet header, downcomer, steam line and inlet feeders. The basic tool for analysis is the TILT-N Code which has been purposely developed for simulating the nuclear, thermal and hydrodynamic behaviour of the CIRENE core and associated heat transport system. An extensive full-scale test programme has been carried out by CNEN and CISE which fully confirms the adequacy of the model. The main results of the analysis show that maximum temperatures are far from those leading to significant fuel damage and that adequate core cooling is provided over the whole transient. (author)

  6. Comparison of methods for the determination of boron in heavy water moderator

    Green, L.W.; Davey, E.C.; Gulens, J.; Longhurst, T.H.; Mislan, J.P.

    1984-01-01

    Five analysis methods were compared for the determination of boron in heavy water moderator: isotope dilution mass spectrometry, spectrophotometry, neutron activation, inductively coupled plasma -atomic emission spectrometry, and ion selective electrode potentiometry. Ten samples were analysed by each method; the results showed close agreement between all of the methods. Only mass spectrometry achieved the required precision ( 10 B concentration must be determined, only mass spectrometry and neutron activation are applicable

  7. Uncertainty analysis of LBLOCA for Advanced Heavy Water Reactor

    Srivastava, A.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

    2008-01-01

    , uncertainty analysis for the Large Break LOCA (200% Inlet Header Break) of Advanced Heavy Water Reactor (AHWR) has been carried out. The uncertainty analysis was carried out for the peak cladding temperature (PCT), based on the two different methods i.e., Wilk's method and the response surface technique. Their findings have also been compared

  8. Parametric studies to establish natural circulation in advanced heavy water reactor

    Bhatia, S K; Dhawan, M L [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Design of Advanced Heavy Water Reactor (AHWR) is in progress. It consists of vertical pressure tubes with boiling light water coolant flowing through the tubes and heavy water moderator in the calandria. In PHWRs, core heat removal is through forced circulation of the coolant by PHT pumps. In AHWR, no PHT pumps are used and core heat is carried away by natural circulation of the coolant due to density difference between steam/water mixture inside the core and the water region outside the core. This passive means of core heat removal results in a number of benefits viz. (a) extra length of piping, valves, instruments, power supply and control systems for functioning of instruments are eliminated, (b) plant layout is simplified, (c) maintenance of valves and instruments is reduced. Natural circulation in AHWR is achieved by keeping the steam drum at a sufficient height above the core to get the required driving force. The loop height depends on many factors i.e. channel power, V{sub c}/V{sub f} ratio (ratio of coolant volume to fuel volume) and core height. The effect of these parameters on the loop height to establish natural circulation have been studied and presented. (author). 1 ref., 1 fig., 1 tab.

  9. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor; Etude des consequences de la rupture d'un tube de force dans la cuve d'un reacteur modere a l'eau lourde et refroidi au gaz

    Hareux, F; Roche, R; Vrillon, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [French] L'eclatement d'un tube de force dans la cuve d'un reacteur de puissance modere a l'eau lourde et refroidi par un gaz sous pression est un accident qui a ete etudie experimentalement a propos d'EL-4. Un premier essai a l'echelle 1 ayant montre que l'eclatement d'un tube ne provoque pas celui des tubes voisins, des essais relatifs a la tenue de la cuve ont ete effectues sur maquettes a echelle tres reduite (l/lO). Il a ete trouve que la cuve peut supporter plusieurs eclatements de tubes sans deformations notables. La pression transitoire dans la cuve a une allure oscillatoire amortie dont le maximum (pression de pic) a fait l'objet d'une etude experimentale detaillee. Il apparait que les parametres essentiels influant sur cette pression sont: la pression du gaz contenu dans le tube de force, le volume du gaz qui participe a l'eclatement, la flexibilite de la cuve, la masse d'air empoisonnee dans la cuve, la nature du gaz explosant. Une methode generale d'estimation des pics de pression dans

  10. Design and development of face seal type sealing plug for advanced heavy water reactor

    Bansal, S.; Bhattacharyya, S.; Patel, R.J.; Agrawal, R.G.; Vaze, K.K.

    2005-09-01

    Advanced Heavy Water Reactor is a vertical pressure tube type reactor having light water as its coolant and heavy water as moderator. Sealing plug is required to close the pressure boundary of main heat transport system of the reactor by preventing escape of light water/steam From the coolant channel. There are 452 coolant channels in the reactor located in square lattice pitch. Sealing plug is located at the top of each coolant channel (in the top end fitting). Top end fitting is having a stepped bore to create a sealing face. Sealing plug is held through its expanded jaws in a specially provided groove of the end fitting. The plug was designed and prototypes were manufactured considering its functional importance, intricate design and precision machining requirements. Sealing plug consists of about 20 components mostly made up of precipitation hardening stainless steel, which is suitable for water environment and meets other requirements of strength and resistance to wear and galling. Seal disc is a critical component of the sealing plug as it is the pressure-retaining component. It is a circular disc with protruded stem. One face of the seal disc is nickel plated in the peripheral area that creates the sealing by abutting against the sealing face provided in the end fitting. The typical shape and profile of seal disc provides flexibility and allows elastic deformation to assist in locking of sealing plug and creating adequate seating force for effective sealing. Design and development aspects of the sealing plug have been detailed out in this report. Also results of stress analysis and experimental studies for seal disc have been mentioned in the report. Stress analysis and experimental testing was required for the seal disc because high stresses are developed due to its exposure to high pressure and temperature environment of Main Heat Transport system. Hot testing was carried out to simulate the reactor-simulated condition. The performance was found to be

  11. Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

    1999-11-01

    In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

  12. Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2000-01-01

    Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

  13. Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

  14. Aspects on thorium utilization in heavy water reactors

    1978-11-01

    Some of the main problems of the Th - PHWR cycles are analyzed. With respect to the burnup limitations introduced by SSET cycle conditions and the burnup sensitivities of this reference cycle, estimates are presented using an integrated neutron-heavy element balance method. A PHWR of 1 GW(e) very similar to the CANDU current design was selected. In the case of 0.5% uranium losses, 11000 MWD/tHE and 13000 MWD/tHE were considered for U-235, respectively, for the Pu initialization of the cycle, the corresponding inventory being 4 t U-235/GW(e) and 5 t Pu (with 72% fissile content) per GW(e) for one year delay time between reactor out to reactor in, 66% capacity factor, 27 MW (fission)/tHE medium specific power. The following aspects are also analyzed: Safety problems associated with low delayed neutron fraction values; High and intermediate burnup fuel elements conceptual problems; Specific problems of thorium reprocessing; Specific problems for radioactive wastes and thorium storage; U-232 content evaluations and related fuel fabrication problems

  15. Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

    Aas, S.; Jamne, E.; Wullum, T.; Fjellestad, K. [Institutt for Atomenergi, OECD Halden Reactor Project, Halden (Norway)

    1968-04-15

    The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm{sup 2}. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO{sub 2} : the most highly exposed elements have reached 10000 MWd/t UO{sub 2}. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 {mu}C/cm{sup 3}. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel

  16. Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis

    Downar, T.J.

    1987-01-01

    A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback

  17. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    Roche, R.; Gaudez, J.C.

    1964-01-01

    In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

  18. The variation of the reactivity with the number, diameter and length of the control rods in a heavy water natural uranium reactor

    McCriric, H

    1958-05-15

    Starting with the known reactor constants for a heavy water moderated reactor with reflector and a given number of control rods of a certain size, the reactivity equivalence of the control rods is calculated. The calculation is given in detail. The number, length and diameter of the control rods is then varied and the effect of these parameters on the reactivity is shown graphically. Flux plots are also given for the reactor with and without control rods.

  19. Considerations regarding design of ion exchange columns for applications in heavy water nuclear reactors- a comprehensive review

    Joginder Kumar; Nema, M.K.

    2000-01-01

    In nuclear reactor applications the principal role of the purification system is to maintain a satisfactory chemistry of moderator and coolant which are different at various stages of reactor operations e.g. during reactor start up, for removal of neutron poison from the moderator, the purification flows are much different compared to steady state operation of the reactor. In order to cater to varying requirements regarding purification load, optimisation in connection with ion exchange column design plays an important role and becomes very challenging in Heavy Water Nuclear Reactors mainly due to the fact that heavy water is very very expensive. In this paper a comprehensive review is made for various designs adopted so far regarding IX column in Indian PHWRs of 220 MWe size for normal operations. Design and operating experience regarding large size IX column used for occasional needs during dilute chemical decontamination of 220 MWe PHWRs is also discussed. The experience regarding development testing of the proposed design of ion exchange column for 500 MWe PHWRs is also discussed

  20. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    Martin, R.; Roche, R.

    1964-01-01

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [fr

  1. Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

    Okubo, T.; Iwamura, T.; Takeda, R.; Yamamoto, K.; Okada, H.

    2001-01-01

    An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

  2. Analysis of severe accidents in pressurized heavy water reactors

    2008-06-01

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

  3. Study relating to the physico-chemical behaviour of heavy water in nuclear reactors

    Chenouard, J.; Dirian, G.; Roth, E.; Vignet, P.; Platzer, R.

    1959-01-01

    Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [fr

  4. Feasibility study and economic analysis on thorium utilization in heavy water reactors

    1978-07-01

    Even though natural uranium is a more easily usable fuel in heavy water reactors, thorium fuel cycles have also been considered owing to certain attractive features of the thorium fuel cycle in heavy water reactors. The relatively higher fission neutron yield per thermal neutron absorption in 233 U combined with the very low neutron absorption cross section of heavy water make it possible to achieve breeding in a heavy water reactor operating on Th- 233 U fuel cycle. Even if the breeding ratio is very low, once a self-sustaining cycle is achieved, thereafter dependence on uranium can be completely eliminated. Thus, with a self-sustaining Th- 233 U fuel cycle in heavy water reactors, a given quantity of natural uranium will be capable of supporting a much larger installed generating capacity to significantly longer period of time. However, since thorium does not contain any fissile isotope, fissile material has to be added at the beginning. Concentrated fissile material is considerably more expensive than the 235 U contained in natural uranium. This makes the fuel cycle cost higher with thorium fuel cycle, at least during the initial stages. The situation is made worse by the fact that, because of its higher thermal neutron absorption cross section, thorium requires a higher concentration of fissile material than 238 U. Nevertheless, because of the superior nuclear characteristics of 233 U, once uranium becomes more expensive, thorium fuel cycle in heavy water reactors may become economically acceptable. Furthermore, the energy that can be made available from a given quantity of uranium is considerably increased with a self-sustaining thorium fuel cycle

  5. Deuterium and heavy water

    Vasaru, G.; Ursu, D.; Mihaila, A.; Szentgyorgyi, P.

    1975-01-01

    This bibliography on deuterium and heavy water contains 3763 references (1932-1974) from 43 sources of information. An author index and a subject index are given. The latter contains a list of 136 subjects, arranged in 13 main topics: abundance of deuterium , catalysts, catalytic exchange, chemical equilibria, chemical kinetics, deuterium and heavy water analysis, deuterium and heavy water properties, deuterium and heavy water separation, exchange reactions, general review, heavy water as moderator, isotope effects, synthesis of deuterium compounds

  6. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

    1995-01-01

    Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

  7. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Owen, M.B.

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR

  8. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  9. Experimental determination of lattice parameters for 2% enriched uranium heavy water reactor cores

    Raisic, N; Takac, S; Markovic, H; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Systematic measurements of the buckling, infinite multiplication factor and the thermal utilization factor were made on a series of lattices for 2% enriched uranium tubular fuel elements in heavy water. This work represents the first phase of experimental verification of standard theoretical methods used for the determination of reactor parameters.

  10. Neutronic study of the two french heavy water reactors

    Horowitz, J.

    1955-01-01

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

  11. Summary report of the 7th reduced-moderation water reactor workshop

    Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

    2005-08-01

    As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

  12. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  13. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  14. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  15. Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

    Thakur, A.; Singh, B.; Pushpam, N.P.; Bharti, V.; Kannan, U.; Krishnani, P.D.; Sinha, R.K.

    2011-01-01

    The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

  16. Study of the light emitted in the moderation of a heavy-water pile

    Breton, D.

    1958-06-01

    During the running of a reactor which uses water as a neutron moderator, a bluish light is seen to appear inside the liquid. A detailed study of this radiation, undertaken on the Fontenay-aux-Roses pile, has shown that the spectrum is identical with that which characterises the light produced by the Cerenkov effect. The light intensity as a function of the pile power grows exponentially as a function of time when the pile diverges, with a lifetime equal to that of the rise in power. An examination of the various particles present in the pile has led to the conclusion that only electrons with an energy greater than 260 keV con produce the Cerenkov light. The light source thus produced is about 2.10 6 photons/cm 2 of water, when the pile power equals 1 watt. (author) [fr

  17. Drying of heavy water system and works of charging heavy water in Fugen

    Matsushita, Tadashi; Iijima, Setsuo

    1980-01-01

    The advanced thermal reactor ''Fugen'' is the first heavy water-moderated, boiling light water-cooled nuclear reactor for power generation in Japan. It is a large heavy water reactor having about 130 m 3 of heavy water inventory and about 300 m 3 of helium space as the cover gas of the heavy water system. The heavy water required was purchased from FRG, which had been used for the power output test in the KKN, and the quality was 99.82 mol % mean heavy water concentration. The concentration of heavy water for Fugen used for the nuclear design is 99.70 mol%, and it was investigated how heavy water can be charged without lowering the concentration. The matters of investigation include the method of bringing the heavy water and helium system to perfect dryness after washing and light water test, the method of confirming the sufficient dryness to prevent the deterioration, and the method of charging heavy water safely from its containers. On the basis of the results of investigation, the actual works were started. The works of drying the heavy water and helium system by vacuum drying, the works of sampling heavy water and the result of the degree of deterioration, and the works of charging heavy water and the measures to the heavy water remaing in the containers are described. All the works were completed safely and smoothly. (J.P.N.)

  18. Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2

    Nikolic, M.

    1963-01-01

    It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events [sr

  19. Plant life management strategies for pressurized heavy water reactors

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  20. Enriched uranium cycles in pressurized heavy water reactors

    Mazzola, A.

    1994-01-01

    A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

  1. Mathematical model for safety analysis of heavy water power reactor

    Milovanovic, M.; Humo, E.; Mitrovic, S.

    1966-01-01

    Fundamental information in formulating the mathematical model for accident analysis is concerned with reactivity changes of the system. These parameters are: changes of fuel and moderator temperature, changes of the upper reflector thickness, reactivity changes due to moderator density variation dependent on the steam quantity and neutron flux distribution in the core

  2. Reactivity effect of a heavy water tank as reflector in the IPEN/MB-01 reactor

    Santos, Adimir dos; Fuga, Rinaldo

    2013-01-01

    This experiment comprises a set of experiments performed in the IPEN/MB-01 reactor and described in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, specifically the experiment aim to evaluate the reactivity due to the heavy water tank placed at reflector region of the IPEN/MB-01 reactor. An aluminum tank was designed to be filled with heavy water and positioned at the west face of the IPEN/MB-01, additionally the experiment was also designed to allow variable heavy water height inside of this tank providing different neutron leakage rate in the west face of the IPEN/MB-01, consequently providing a series of interesting combinations. The measured quantities in the experiment are reactivities and critical control bank positions for several combinations of the control banks and an excess of reactivity of the heavy water tank. The experiment will be simulated using a Monte Carlo code MCNP in order to compare the different critical control bank position. (author)

  3. Preliminary Assessment of Heavy-Water Thorium Reactors in the Brazilian Nuclear Programme

    Salvo Brito, S. de; Lepecki, W. P.S. [Instituto de Pesquisas Radioativas, Belo Horizonte (Brazil)

    1968-04-15

    Since December 1965, the Instituto de Pesquisas Radioativas has been studying for the Brazilian Nuclear Energy Commission the feasibility of a thorium reactor programme in Brazil; since June 1966, the programme has been developed in close co-operation with the French Atomic Energy Commission. A reference conceptual design of a heavy-water-cooled and -moderated thorium converter reactor has been developed. The main features of that concept are the use of a prestressed-concrete pressure vessel, integrated arrangement of the primary circuit and the possibility of on-load fuel management. Economic competitiveness could be the result of high compactness, low capital costs and low fuel consumption. The technology involved is not very sophisticated; intensive engineering development work must be done in areas like fuel charge machine, concrete vessel insulation, and proper design of heat exchangers, but it is the feeling of the Group that these problems could be solved without seriously compromising the economic feasibility of the concept. Preliminary studies were made on the alternative use of enriched uranium or plutonium as a feed for the programme; in the latter case, plutonium could be produced in natural uranium reactors of the same type. The general conditions favouring each of these approaches to the thorium cycle have been determined, in particular those related to the costs of the fissile materials in the world market and to the country's policy related to nuclear fuel imports. The results of the preliminary studies are very encouraging and could justify the beginning of a research and development programme leading to the construction of a prototype in the 1970's. (author)

  4. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  5. Status of research and development on reduced-moderation water reactors

    Iwamura, Takamichi

    2002-01-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  6. Status of research and development on reduced-moderation water reactors

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  7. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  8. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

  9. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  10. A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

    Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

  11. Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

    Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

    2017-02-15

    The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

  12. Investigation of the heavy water distillation system at the RA reactor; Ispitivanje sistema za destilaciju teske vode reaktora RA

    Zecevic, V; Badrljica, R [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    The heavy water distillation system was tested because this was not done before the reactor start-up. Detailed inspection of the system components showed satisfactory results. Leak testing was done as well as the testing of the instrumentation which enables reliable performance of the system. Performance testing was done with ordinary water and later 2700 l of heavy water from the reactor was purified, decreasing the activity by 45%.

  13. Benchmark testing of Canadol-2.1 for heavy water reactor

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  14. Development of a silicon calorimeter for dosimetry applications in a water-moderated reactor

    DePriest, Kendall Russell; King, Donald Bryan; Naranjo, Gerald E.; Luker, Spencer Michael; Keltner, Ned R.; Suo-Anttila, Ahti Jorma; Griffin, Patrick Joseph

    2005-01-01

    High fidelity active dosimetry in the mixed neutron/gamma field of a research reactor is a very complex issue. For passive dosimetry applications, the use of activation foils addresses the neutron environment while the use of low neutron response CaF 2 :Mn thermoluminescent dosimeters (TLDs) addresses the gamma environment. While radiation-hardened diamond photoconducting detectors (PCD) have been developed that provide a very precise fast response (picosecond) dosimeter and can provide a time-dependent profile for the radiation environment, the mixed field response of the PCD is still uncertain and this interferes with the calibration of the PCD response. In order to address the research reactor experimenter's need for a dosimeter that reports silicon dose and dose rate at a test location during a pulsed reactor operation, a silicon calorimeter has been developed. This dosimeter can be used by itself to provide a dose in rad(Si) up to a point in a reactor pulsed operation, or, in conjunction with the diamond PCD, to provide a dose rate. This paper reports on the development, testing, and validation of this silicon calorimeter for applications in water-moderated research reactors.

  15. Expansion connection of socket in flow distributed cabin of heavy water research reactor inner shell

    Jiang Zhiliang; Li Yanshui

    1995-01-01

    Expansion connection of aluminium alloy LT21 socket in flow distributed cabin of Heavy Water Research Reactor (HWRR) inner shell is described systematically. The expansion connection technology parameters of products are determined through tests. They are as following: bounce value of inner diameter after expansion, expansion degree, space between socket and plate hole, device for expanding pipes, selection of tools for enlarging or reaming holes, manufacture for socket inner hole and cleaning after expansion

  16. HAZOP-study on heavy water research reactor primary cooling system

    Hashemi-Tilehnoee, M.; Pazirandeh, A.; Tashakor, S.

    2010-01-01

    By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.

  17. The status of improved pressurized heavy water reactor development - A new option for PHWR -

    Park, Tae Keun; Yeo, Ji Won

    1996-03-01

    Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

  18. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    Shunsuke Uchida; Eishi Ibe; Katsumi Ohsumi

    1994-01-01

    Application of hydrogen water chemistry to moderate corrosive circumstances is a promising approach to preserve structural integrities of major components and structures in the primary cooling system of BWRs. The benefits of HWC application are usually accompanied by several disadvantages. After evaluating merits and demerits of HWC application, it is concluded that optimal amounts of hydrogen injected into the feed water can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. (authors). 1 fig., 4 refs

  19. Implications of alpha-decay for long term storage of advanced heavy water reactor fuels

    Pencer, J.; McDonald, M.H.; Roubtsov, D.; Edwards, G.W.R.

    2017-01-01

    Highlights: •Alpha decays versus storage time are calculated for examples of advanced heavy water reactor fuels. •Estimates are made for fuel swelling and helium bubble formation as a function of time. •These predictions are compared to predictions for natural uranium fuel. •Higher rates of damage are predicted for advanced heavy water reactor fuels than natural uranium. -- Abstract: The decay of actinides such as 238 Pu, results in recoil damage and helium production in spent nuclear fuels. The extent of the damage depends on storage time and spent fuel composition and has implications for the integrity of the fuels. Some advanced nuclear fuels intended for use in pressurized heavy water pressure tube reactors have high initial plutonium content and are anticipated to exhibit swelling and embrittlement, and to accumulate helium bubbles over storage times as short as hundreds of years. Calculations are performed to provide estimates of helium production and fuel swelling associated with alpha decay as a function of storage time. Significant differences are observed between predicted aging characteristics of natural uranium and the advanced fuels, including increased helium concentrations and accelerated fuel swelling in the latter. Implications of these observations for long term storage of advanced fuels are discussed.

  20. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    2013-12-01

    Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

  1. Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

    Sankaranarayanan, S.

    2004-01-01

    Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

  2. Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

    Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

    2005-01-01

    R and D project to investigate thermal-hydraulic performance of tight-lattice fuel bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in collaboration with utilities, reactor vendors and universities from 2002. The RMWR realizes a high conversion ratio larger than 0.1 for sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The reactor core comprises tight-lattice fuel assemblies with gap clearance of around 1.0 mm to reduce the water volume ratio to achieve the high conversion ratio. A problem of utmost importance from a thermal-hydraulic point of view is the coolability of the tight-lattice assembly with such a small gap width. JAERI has been carrying out experimental study to investigate the system parameter effects on the thermal-hydraulic performance and to confirm the feasibility of the core. In the present study, the subchannel analysis code NASCA was applied to 37-rod tight-lattice bundle experiments. The NASCA can give good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy decreases for the gap width of 1.0 mm. To improve the prediction accuracy, the code will be modified to take the effect of film thickness distribution around fuel rods on boiling transition. (author)

  3. Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

    Mosteller, R.D.; Hall, R.A.; Lancaster, D.B.; Young, E.H.; Gavin, P.H.; Robertson, S.T.

    1998-01-01

    The contents of ANS 19.11, the standard for ''Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,'' are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard

  4. Sources of variability for the single-comparator method in a heavy-water reactor

    Damsgaard, E.; Heydorn, K.

    1978-11-01

    The well thermalized flux in the heavy-water-moderated DR 3 reactor at Risoe prompted us to investigate to what extent a single comparator could be used for multi-element determination instead of multiple comparators. The reliability of the single-comparator method is limited by the thermal-to-epithermal ratio, and experiments were designed to determine the variations in this ratio throughout a reactor operating period (4 weeks including a shut-down period of 4-5 days). The bi-isotopic method using zirconium as monitor was chosen, because 94 Zr and 96 Zr exhibit a large difference in their Isub(o)/Σsub(th) values, and would permit determination of the flux ratio with a precision sufficient to determine variations. One of the irradiation facilities comprises a rotating magazine with 3 channels, each of which can hold five aluminium cans. In this rig, five cans, each holding a polyvial with 1 ml of aqueous zirconium solution were irradiated simultaneously in one channel. Irradiations were carried out in the first and the third week of 4 periods. In another facility consisting of a pneumatic tube system, two samples were simultaneously irradiated on top of each other in a polyethylene rabbit. Experiments were carried out once a week for 4 periods. All samples were counted on a Ge(Li)-detector for 95 Zr, 97 sup(m)Nb and 97 Nb. The thermal-to-epithermal flux ratio was calculated from the induced activity, the nuclear data for the two zirconium isotopes and the detector efficiency. By analysis of variance the total variation of the flux ratio was separated into a random variation between reactor periods, and systematic differences between the positions, as well as the weeks in the operating period. If the variations are in statistical control, the error resulting from use of the single-comparator method in multi-element determination can be estimated for any combination of irradiation position and day in the operating period. With the measure flux ratio variations in DR

  5. Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

    Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

    2003-01-01

    The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

  6. Detection of heavy-water leaks in nuclear reactors : a novel method

    Murthy, M.S.; Gor, M.K.

    2002-01-01

    Technical Physics and Prototype Engineering Division, BARC has designed, developed and produced several high sensitivity mass spectrometer helium leak detectors over a period of two decades. Sometimes back, when there was a problem of detecting heavy water leaks in situ in one of the nuclear power reactors of the Department of Atomic Energy, it was referred to this division for a technical solution. After discussing with the site engineers, the various problems involved in the on-line detection of heavy water leaks especially near the end fittings of the coolant assemblies, a novel method of leak detection was developed. Some of the salient features of the method and the results obtained in the laboratory tests are given in this paper. (author)

  7. The supply of steam from Candu reactors for heavy water production

    Robertson, R.F.S.

    1975-09-01

    By 1980, Canada's energy needs for D 2 O production will be 420 MW of electrical energy and 3600 MW of thermal energy (as steam). The nature of the process demands that this energy supply be exceptionally stable. Today, production plants are located at or close to nuclear electricity generating sites where advantage can be taken of the low cost of both the electricity and steam produced by nuclear reactors. Reliability of energy supply is achieved by dividing the load between the multiple units which comprise the sites. The present and proposed means of energy supply to the production sites at the Bruce Heavy Water Plant in Ontario and the La Prade Heavy Water Plant in Quebec are described. (author)

  8. CANDU - Canadian experience and expectations with the heavy-water reactor

    Foster, J.S.; Russell, S.H.

    1977-05-01

    The paper describes the evolution of the CANDU nuclear-power plants with particular reference to the objectives of safety, reliability and economy; the development of industrial capacity for the supply of fuel, components and heavy water; and the prospective development of advanced fuel cycles and the projected results. It provides data on radiation, releases, and exposures, internal and external to the power plants; plant availability, capacity factors and other performance data; heavy water production data with reference to safety, reliability, and economics; projections of the performance of CANDU reactors operating on a thorium-U-233 cycle and the development required to establish this cycle; and intent with respct to spent-fuel management and radioactive-waste storage. (author)

  9. Experience with dilute chemical decontamination in Indian Pressurized Heavy Water Reactors

    Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Subramanian, Veena; Mittal, V.K.; Narasimhan, S.V.

    2010-01-01

    Dilute Chemical Decontamination (DCD) process has been used in several full system and components of nuclear coolant systems to effectively remove the radioactive contaminants that causes radiation field and consequent MANREM problem. The DCD process uses chemicals in very low concentrations (millimolar) and dissolves the oxide film along with the activity incorporated in the oxide film. In DCD process operated under the regenerative mode, the chemical formulation spent in the process of oxide dissolution is replenished by passing through cation exchange columns. Finally, after achieving sufficient decontamination of the system/component, the added decontamination chemicals along with the activities and metal ions released during the process are removed by mixed bed ion exchange columns and the system is restored to normal operating condition in few days time. In PHWRs, the regenerative DCD process is applied for full primary coolant system decontamination. The chemicals are added directly to the heavy water coolant with the fuel in the core. In Indian PHWRs (MAPS-1 and 2, RAPS-1 and 2, NAPS-1 and 2 and KAPS-1), the process has been applied eleven times. A chemical formulation based on NTA, Citric acid and Ascorbic acid has been applied seven times with good results. Decontamination factors in the range 2-30 have been obtained in different components with good MANREM savings in the subsequent maintenance works. Efforts are on to modify the process to take care of the challenges posed by antimony isotope. An inhibitor (Rodine-92B) based process was successfully tested in NAPS-2 for removing antimony isotopes ( 122 Sb and 124 Sb). Further refining of the antimony removal process is being worked out. Similarly, the process is being modified to effectively remove the hotspot causing stellite particles in the moderator system of PHWRs. A permanganate based process has been developed and tested in several adjustor rod drive mechanisms in KAPS and NAPS. The experience of

  10. Conceptual design of a quasi-homogeneous pressurized heavy water reactor to be operated in the closed Th-U233 fuel cycle

    1979-06-01

    This paper deals with the heavy water reactor, which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources not wishing to be dependent on the import of enrichment and other fuel cycle services which, in addition, would draw on the foreign exchange reserves. Pressurized heavy water reactors have been designed and built according to two somewhat different versions. While the Canadian CANDU-PHWR concept uses pressure tubes in a nearly unpressurized moderator tank (calandria), the German development line takes advantage of the established and well proven LWR technology, and, thus, uses a pressure vessel design where coolant channels and the surrounding moderator are held at equal pressure. This pressure vessel type heavy water reactor which has been built on a commercial demonstration plant level at ATUCHA in Argentina is described in a companion paper where also a conceptual design for a 685 MWsub(e) PHWR is discussed

  11. All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

    2009-01-01

    Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

  12. Energetics of semi-catalyzed-deuterium, light-water-moderated, fusion-fission toroidal reactors

    Jassby, D.L.; Towner, H.H.; Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.

    1978-07-01

    The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U 3 Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10 14 cm -3 s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor

  13. Natural uranium fueled light water moderated breeding hybrid power reactors: a feasibility study

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    1978-06-01

    The first part of the study consists of a thorough investigation of the properties of subcritical thermal lattices for hybrid reactor applications. Light water is found to be the best moderator for (fuel-self-sufficient) FSS hybrid reactors for power generation. Several lattice geometries and compositions of particular promise for LWHRs are identified. Using one of these lattices, fueled with natural uranium, the performance of several concepts of LWHR blankets is investigated, and optimal blanket designs are identified. The effect of blanket coverage efficiency and the feasibility of separating the functions of tritium breeding and of power generation to different blankets are investigated. Optimal iron-water shields for LWHRs are also determined. The performance of generic types of LWHRs is evaluated. The evolution of the blanket properties with burnup is evaluated and fuel management schemes are briefly examined. The feasibility of using the lithium system of the blanket to control the blanket power amplitude and shape is also investigated. A parametric study of the energy balance of LWHR power plants is carried out, and performance parameters expected from LWHRs are estimated. Discussions are given of special features of LWHRs and their fuel cycle

  14. Pressure drop characteristics in tight-lattice bundles for reduced-moderation water reactors

    Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

    2004-01-01

    The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4 x 5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3 mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier. (author)

  15. Indian heavy water programme - challenges and opportunities

    Aruldoss Kanthiah, W.S.

    2010-01-01

    Discovery of fission of uranium in 1939 opened up hitherto unknown possibilities for utilising the fission energy for use of mankind, mainly for the production of and electrical energy. It was realised that this nuclear energy could be an ideal substitute for the fast depleting fossil fuels which would one day get exhausted. Two main concepts of nuclear power reactor got evolved, one enriched uranium fuelled, ordinary water moderated reactor and another natural uranium fuelled heavy water moderated reactor. The concentration of uranium 235 U needed for ordinary water moderated reactors is 3% but the naturally occurring uranium in India contains only 0.7% of 235 U. The reactors utilising natural uranium as fuel require Heavy Water as moderator. The processing of uranium ore to achieve from 0.7% to 3% is highly complex. Recognising the fact that India has limited uranium resources but rich thorium resources, Dr. Bhabha formulated a three stage nuclear power generation programme for our country. The first generation reactors can use natural uranium as fuel with heavy water as moderator. Since the technology to generate such large scale heavy water to match the urgent need for nuclear power generation was not indigenously available, the technology available with Canada and France was utilised for installation of first generation heavy water plants in India. However, the peaceful nuclear experiment conducted by India in 1974 caused resentment among the countries that supplied Heavy Water technology to India and they stopped all technological help and assistance in nuclear field. Thereafter, it was the story of India going alone in heavy water production. That made India meets successfully all challenges on the way to installation, commissioning and sustained operation of all plants. Today we have six operating Heavy Water plants, spread all over the country. We have reached a stage, a change from a situation of crunch to a level of not only self sufficiency but to a

  16. The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

    Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

    1992-01-01

    A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

  17. Analytical modelling and study of the stability characteristics of the Advanced Heavy Water Reactor

    Nayak, A.K.; Vijayan, P.K.; Saha, D.

    2000-04-01

    An analytical model has been developed to study the thermohydraulic and neutronic-coupled density-wave instability in the Indian Advanced Heavy Water Reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the design configuration considered may experience both Ledinegg and Type I and Type II density-wave instabilities depending on the operating condition. Some methods of suppressing these instabilities were found out. In addition, it was found that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have strong influence on the Type I and Type II instabilities. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (author)

  18. Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

    Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

    2004-06-01

    The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

  19. Nuclear power reactors

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  20. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  1. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  2. Control of xenon oscillations in Advanced Heavy Water Reactor via two-stage decomposition

    Munje, R.K.; Parkhe, J.G.; Patre, B.M.

    2015-01-01

    Highlights: • Singularly perturbed model of Advanced Heavy Water Reactor is explored. • Composite controller is designed using slow subsystem alone, which achieves asymptotic stability. • Nonlinear simulations are carried out under different transient conditions. • Performance of the controller is found to be satisfactory. - Abstract: Xenon induced spatial oscillations developed in large nuclear reactors, like Advanced Heavy Water Reactor (AHWR) need to be controlled for safe operation. Otherwise, a serious situation may arise in which different regions of the core may undergo variations in neutron flux in opposite phase. If these oscillations are left uncontrolled, the power density and rate of change of power at some locations in the reactor core may exceed their respective thermal limits, resulting in fuel failure. In this paper, a state feedback based control strategy is investigated for spatial control of AHWR. The nonlinear model of AHWR including xenon and iodine dynamics is characterized by 90 states, 5 inputs and 18 outputs. The linear model of AHWR, obtained by linearizing the nonlinear equations is found to be highly ill-conditioned. This higher order model of AHWR is first decomposed into two comparatively lower order subsystems, namely, 73rd order ‘slow’ subsystem and 17th order ‘fast’ subsystem using two-stage decomposition. Composite control law is then derived from individual subsystem feedback controls and applied to the vectorized nonlinear model of AHWR. Through the dynamic simulations it is observed that the controller is able to suppress xenon induced spatial oscillations developed in AHWR and the overall performance is found to be satisfactory

  3. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Jayalal, M.L.; Ramachandran, Suja; Rathakrishnan, S.; Satya Murty, S.A.V.; Sai Baba, M.

    2015-01-01

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  4. From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca

    Stefanovic, D.; Pesic, M.

    1995-01-01

    A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

  5. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

    Leszczynski, Francisco

    2000-01-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

  6. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  7. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  8. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  9. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  10. Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

    High, H.M.

    1983-05-01

    The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

  11. Removal of decay heat by specially designed isolation condensers for advanced heavy water reactors

    Dhawan, M L; Bhatia, S K [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    For Advanced Heavy Water Reactor (AHWR), removal of decay heat and containment heat is being considered by passive means. For this, special type of isolation condensers are designed. Isolation condensers when submerged in a pool of water, are the best choice because condensation of high temperature steam is an extremely efficient heat transfer mechanism. By the use of isolation condensers, not only heat is removed but also pressure and temperature of the system are automatically controlled without losing the coolant and without using conventional safety relief valves. In this paper, design optimisation studies of isolation condensers of different types with natural circulation for the removal of core decay heat for AHWR is presented. (author). 8 refs., 2 figs.

  12. Radiation characteristics of spent fuel of heavy-water research reactor during long-term storage

    Gerasimov, A.S.; Kiselev, G.V.; Myrtsymova, L.A.; Zaritskaya, T.S.

    2002-01-01

    Decay heat power and radiotoxicity by water of actinides and fission products from spent fuel of heavy-water research reactor RA were calculated for period of storage during 300000 years. Three variants of fuel enrichment by 235 U were considered: 2%, 21%, and 80%. The mass of 235 U in one fuel element was supposed to be the same for all variants of enrichment. The decay heat power of fission products in initial period is about 20 times higher than that of actinides. Decay heat power and radiotoxicity of actinides do not practically decrease during long period of time as they are determined by nuclides with very long half-life periods. (author)

  13. Acoustic emission technique for leak detection in an end shield of a pressurized heavy water reactor

    Kalyanasundaram, P.; Jayakumar, T.; Raj, B.

    1989-01-01

    This paper discusses the successful application of the Acoustic Emission Technique (AET) for detection and location of leak paths present on the inaccessible side of an end shield of a Pressurized Heavy Water Reactor (PHWR). The methodology was based on the fact that air and water leak AE signals have different characteristic features. Baseline data was generated from a sound end-shield of a PHWR for characterizing the background noise. A mock up end-shield system with saw cut leak paths was used to verify the validity of the methodology. It was found that air leak signals under pressurisation (as low as 3 psi) could be detected by frequency domain analysis. Signals due to air leaks from various locations of a defective end-shield were acquired and analysed. It was possible to detect and locate leak paths. Presence of detected leak paths were further confirmed by alternate test. (orig.)

  14. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

    1996-01-01

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

  15. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

    1997-12-31

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

  16. Design guide for category II reactors light and heavy water cooled reactors

    Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems

  17. Safety analysis of high temperature reactor cooled and moderated by supercritical light water

    Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

    2003-01-01

    This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

  18. Development of long-life neutron detectors for the prototype heavy water reactor 'Fugen'

    Ohteru, Shigeru; Shirayama, Shimpey.

    1981-01-01

    The development of long-life neutron detectors as the flux monitors for the prototype heavy water reactor has been made. Three kinds of neutron monitors, namely start-up monitor (SUM), power up monitor (PUM) and local power monitor (LPM), are provided. The LPM consists of 4 ion chamber type neutron detectors and a guide tube of power calibration monitor (PCM). This is useful for reactor control and fuel soundness monitor. The improvement of the neutron detectors was made for the operation under high neutron flux and gamma-ray heating. For the long-life operation, U-234 was mixed into U-235 for the conversion in the detectors. The ratio of U-234 to U-235 is 3 to 1. The PCM is also an ion chamber type detector with U-235. The mixing ratio of U-234 to U-235 was determined by a test with the JMTR. The characteristic performance was also investigated by the JMTR. After the completion of Fugen, various tests on the long-life detectors were performed with Fugen. It was hard to test the output linearity of the detectors with a large scale reactor. Therefore, it was tested that the operation range of the detectors is within the linear region of detector output. The voltage-current characteristics and the correlation of output current and saturation current were measured. The variation of the neutron sensitivity of the detectors with the cumulative dose was also studied. (Kato, T.)

  19. A Conceptual Supercritical Water Cooled Reactor Design Using a Cruciform Solid Moderator

    Joo, Hyung Kook; Bae, Kang Mok; Yoo, Jae Woon; Lee, Hyun Chul; Noh, Jae Man; Bae, Yoon Yong

    2006-01-15

    A Super Critical Water-Cooled Reactor(SCWR) concept proposed by Gen-IV has an advantage of a high thermal efficiency. However, there are some difficulties in neutronic core design for a SCWR due to lower moderator density resulting from the high operating temperature over the pseudo-critical temperature. In this report, the design concepts for the fuel assembly and the core for a SCWR were described as a feasibility study on the SCWR core design. HELIOS lattice code which will be used for group constants generation was verified for the application to the low coolant density condition of a SCWR. The TAF module for a thermal hydraulic feedback in MASTER was modified to consider high pressure and temperature of the supercritical coolant with single-phase fluid. A cruciform ZrH{sub 2} solid moderator was proposed for the SCWR fuel assembly design to compensate the lower coolant density. The axial zoning concept with three different enrichments for a fuel rod was used for the axial power shape control. Gadolinia burnable poison rods were used to reduce excess reactivity. Control rod system was grouped into 6 banks to control the excess reactivity of the core during normal operation. An orifice concept for each assembly was applied to control a coolant flow rate individually. As a result of the neutronic analysis for the equilibrium SCWR core, the maximum linear heat generation rete limit was satisfied and the maximum coolant temperature of the core outlet was {approx}590 .deg. C which is lower than 620 .deg. C of the maximum clad temperature limit.

  20. Thermophysical properties database of materials for light water reactors and heavy water reactors. Final report of a coordinated research project 1999-2005

    2006-06-01

    The IAEA Coordinated Research Project (CRP) on the Establishment of a Thermo-physical Properties Database for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs) started in 1999. It was included in the IAEA's Nuclear Power Programme following endorsement in 1997 by the IAEA's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and the TWG-HWR). Furthermore, the TWG on Fuel Performance and Technology (TWG-FPT) also expressed its support. This CRP was conducted as a joint task within the IAEA's project on technology development for LWRs and HWRs in its nuclear power programme. Improving the technology for nuclear reactors through better computer codes and more accurate materials property data can contribute to improved economics of future plants by helping to remove the need for large design margins, which are currently used to account for limitations of data and methods. Accurate representations of thermo-physical properties under relevant temperature and neutron fluence conditions are necessary for evaluating reactor performance under normal operation and accident conditions. The objective of this CRP was to collect and systematize a thermo-physical properties database for light and heavy water reactor materials under normal operating, transient and accident conditions and to foster the exchange of non-proprietary information on thermo-physical properties of LWR and HWR materials. An internationally available, peer reviewed database of properties at normal and severe accident conditions has been established on the Internet. This report is intended to serve as a useful source of information on thermo-physical properties data for water cooled reactor analyses. The properties data have been initially stored in the THERSYST data system at the University of Stuttgart, Germany, which was subsequently developed into an internationally available Internet database named THERPRO at Hanyang University, Republic of Korea

  1. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    Fukaya, Y.; Okubo, T.; Uchikawa, S.

    2008-01-01

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

  2. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2008-07-15

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

  3. Evaluation of a dilute chemical decontaminant for pressurized heavy water reactors

    Velmurugan, S.; Narasimhan, S.V.; Mathur, P.K.; Venkateswarlu, K.S.

    1991-01-01

    In this paper a dilute chemical decontamination formulation based on ethylene diamine tetraacetic acid, oxalic acid, and citric acid is evaluated for its efficacy in removing oxide layers in a pressurized heavy water reactor (PHWR). An ion exchange system that is specifically suited for fission product-dominated contamination in a PHWR is suggested for the reagent regeneration stage of the decontamination process. An attempt has been made to understand the redeposition behavior of various isotopes during the decontamination process. The polarographic method of identifying the species formed in the dissolution process is explained. Electrochemical measurements are employed to follow the course of oxide removal during the dissolution process. Scanning electron micrographs of metal coupons before and after the dissolution process exemplify the involvement of base metal in the formation of a ferrous oxalate layer. Material compatibility tests between the decontaminant and carbon steel, Monel-400, and Zircaloy-2 are reported

  4. MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

    Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

    2011-01-01

    In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

  5. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs

  6. Comparison Of The Worth Of Critical And Exponential Measurements For Heavy-Water-Moderated Reactors; Valeur Relative des Mesures Critiques et Exponentielles pour l'Etude des Reacteurs Ralentis a l'Eau Lourde; Sravnenie tsennosti kriticheskikh i ehksponentsial'nykh izmerenij dlya reaktorov s tyazhelovodnym zamedlitelem; Valor Relativo de las Mediciones Criticas y Exponenciales para los Reactores Moderados por Agua Pesada

    Graves, W. E.; Hennelly, E. J. [Savannah River Laboratory, E.I. Du Pont De Nemours and Co., Aiken, SC (United States)

    1964-02-15

    Critical and exponential experiments in general produce overlapping information on reactor lattices. Over the past ten years the Savannah River Laboratory has been operating a heavy-water critical, the PDP, and an exponential, the SE, in parallel. This paper summarizes SRL experience to give results and recommendations as to the applicability of the two kinds of facilities in different experiments. Six types of experiments are considered below: (1) Buckling measurements in uniform isotropic lattices Here Savannah River has made extensive comparisons between single-region criticals, exponentials, substitution criticals, and PCTR type measurements. The only difficulties in the exponentials seem to lie in the radial-buckling determinations. If these can be made successfully, the exponentials can offer good competition to the criticals. Material requirements are greatest for the single-region criticals, roughly comparable for the substitution criticals and exponentials, and least for the PCTR measurements. (2) Anisotropic and void effects SRL experiments with the criticals and with critical-exponential comparisons are reviewed briefly here and at greater length in a companion paper. (3) Evaluation of control systems Adequately analysed exponential experiments appear to give good results for total-worth measurements. However, for adequate study of overall flux shaping, flux tilts, etc. a full-sized critical such as the PDP is required. (4) Temperature coefficients Exponential experiments provide an excellent method for determining the temperature coefficient of buckling for uniform lattice heating. A special facility, the PSE, at Savannah River permits such measurements up to temperatures of 215 Degree-Sign C. For non-uniform lattice heating criticals are generally preferred. (5) Mixed lattices Actual reactors rarely use the simple uniform lattices to which the exponentials basically apply. Critical experiments with mixed loadings are used at SRL both in measuring

  7. Hydrodynamically induced dryout and post dryout important to heavy water reactors: A yearly progress report

    Ishii, M.; Revankar, S.T.; Babelli, I.; Lele, S.

    1992-06-01

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed to study the CBF induced by various flow instabilities. Details of this test loop are presented. Initial test results have been presented. The two-phase flow regimes and hydrodynamic behaviors in the post dryout region have been studied under propagating rewetting conditions. The effect of subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CBF location and hence on the flow regime encountered in the pre-CBF region

  8. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  9. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  10. Determination of the tritium content in the reactor heavy water; Odredjivanje porasta kolicine tritijuma u reaktorskoj teskoj vodi

    Ribnikar, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Quantity of tritium was measured in the heavy water, in the heavy water vapour and radiolyzed deuterium from the helium cover gas of the RA reactor. It has been shown that isotopic equilibrium D{sub 2}O+DT{r_reversible}DTO+D{sub 2} exists and that it is catalyzed by irradiation. Small quantities of ammonium in the reactor cover gas are inhibiting the migration of tritium from the heavy water. Izmerena ja sadrzina tritijuma u tecnoj teskoj vodi, njenoj pari i radiolizovanom deuterijumu iz helijumske atmosfere reaktora RA. Pokazano je da postoji izotopska ravnoteza D{sub 2}O+DT{r_reversible}DTO+D{sub 2}, koja je katalizovana zracenjem. Male kolicine amonijaka reaktorske atmosfere deluju u smislu otezavanja migracije tritijuma iz teske vode (author)

  11. Study of the modifications on the synchronous generators, heavy water pumps and condenser batteries of the RA reactor - Annex 17

    Milosevic, M.

    1964-01-01

    Modifications done on the synchronous generators are related to the emergency power supply system, meaning one of the most important devices responsible for reactor safety. Without reducing the efficiency of the heavy water pumps the improved stability of generators operation was achieved by reducing the possibility of errors and simplifying manipulation. Condensator batteries were improved in order to decrease the leakage currents

  12. Role of research and development in life management programme and upgradation of safety of Indian Pressurised Heavy Water Reactors

    Rupani, B.B.; Vijayan, P.K.; Rama Rao, A.; Sinha, R.K.

    2009-01-01

    At present, India has a fleet of thirteen small size 220 MWe Pressurised Heavy Water Reactors (PHWRs) and two medium size 540 MWe PHWRs. Reactor Engineering Division (RED) of Bhabha Atomic Research Centre (BARC) has pursued multi-faceted Research and Development programmes to support each phase of PHWR i.e. design, construction, commissioning, operation, maintenance, In-Service Inspection, repair and replacement and life extension, This programme is mainly related to life management of coolant channels, development of tooling and techniques for In-service Inspection of coolant channels, development of repair and replacement technology for coolant channels and moderator system, In-house development of technology and equipments like rolled joints to joint dissimilar metals and lancing equipment for steam generator and state-of art diagnostic systems for trouble shooting critical operating systems. The strong R and D support provided in the programme has significantly contributed towards safe operation of PHWRs. This paper gives the highlights of the major activities in above areas with their end uses and capability. (author)

  13. Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2001-01-01

    In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

  14. Application of noise analysis technique for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors

    Shieh, D.J.; Upadhyaya, B.R.; Sweeney, F.J.

    1987-01-01

    A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer

  15. Daily tritium intakes by people living near a heavy-water research reactor facility: dosimetric significance

    Trivedi, A.; Cornett, R.J.; Galeriu, D.; Workman, W.; Brown, R.M.

    1997-02-01

    We have estimated the relative daily intakes of tritiated water (HTO) and organically bound tritium (OBT), and have measured HTO-in-urine, in an adult population residing in the town of Deep River, Ontario, near a heavy-water research reactor facility at Chalk River. The daily intake of elevated levels of atmospheric tritium has been estimated from its concentration in environmental and biological samples, and various food items from a local tritium-monitoring program. Where the available data were inadequate, we used estimates generated by an environmental tritium-transfer model. From these data and estimates, we calculated a total daily tritium intake of about 55 Bq. Of this amount, 2.5 Bq is obtained from OBT-in-diet. Inhalation of HTO-in-air (15 Bq d -1 ) and HTO-in-drinking water (15 Bq d -1 ) accounts for more than half of the HTO intake. Skin absorption of HTO from air and bathing or swimming (for 30 min d -1 ) accounts for another 9 Bq d -1 and 0.1 Bq d -1 , respectively. The remaining intake of HTO is from food as tissue-free water tritium. The International Commission on Radiological Protection's recommended two-compartment metabolic model for tritium predicts an equilibrium body burden of about 900 Bq from HTO (818 Bq) and OBT (83 Bq) in the body, which corresponds to an annual tritium dose of 0.41 μSv. The model-predicted urinary excretion of HTO (∼18 Bq L -1 ) agrees well with measured HTO-in-urine (range, 10-32 Bq L -1 ). The OBT dose contribution to the total tritium dose is about 16%. We conclude that for the people living near the Chalk River research reactor facility, the bulk of the tritium dose is due to HTO intake. (author)

  16. Heavy water reactors: Status and projected development. Part II. Final draft of a report to be published in the IAEA technical reports series. Working material

    2001-01-01

    retains a low cost operational condition; to illustrate the short and medium term potential for design evolution of the heavy water reactor type; to describe the basis of economic competitiveness of the HWR, its resistance to severe cost increases and capability for extensive source localization; and to provide a reference document on HWRs and to help guide the activities of the IWG-HWR. Those organizations developing and operating HWRs recognize the potential for development of this line of reactors, and it is the intent of this document to illustrate that potential. Various countries and organizations in the past have explored a number of variants of heavy water reactors, and there is a desire to continue to explore some of these options in the future. Currently the pressurized heavy water cooled, heavy water moderated design is an economically competitive design and will likely continue to dominate the heavy water reactor type for some time. This document concentrates on heavy water moderated reactors for electricity production. Reactors for district heating and research reactors are not discussed, except where historical multi-purpose use was a rationale for developing the concept

  17. Heavy water reactors: Status and projected development. Part I. Final draft of a report to be published in the IAEA technical reports series. Working material

    2001-01-01

    retains a low cost operational condition; to illustrate the short and medium term potential for design evolution of the heavy water reactor type; to describe the basis of economic competitiveness of the HWR, its resistance to severe cost increases and capability for extensive source localization; and to provide a reference document on HWRs and to help guide the activities of the IWG-HWR. Those organizations developing and operating HWRs recognize the potential for development of this line of reactors, and it is the intent of this document to illustrate that potential. Various countries and organizations in the past have explored a number of variants of heavy water reactors, and there is a desire to continue to explore some of these options in the future. Currently the pressurized heavy water cooled, heavy water moderated design is an economically competitive design and will likely continue to dominate the heavy water reactor type for some time. This document concentrates on heavy water moderated reactors for electricity production. Reactors for district heating and research reactors are not discussed, except where historical multi-purpose use was a rationale for developing the concept

  18. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  19. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

    2013-01-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  20. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  1. Impact of moderator history on physics parameters in pressurized water reactors

    Mosteller, R.D.

    1988-01-01

    The magnitude of differential reactivity effects that result from spectral differences in different portions of a pressurized water (PWR) core is studied, and it is shown that these effects can be correlated very well with the local moderator history. The impact of these differences on physics parameters such as axial offset, isothermal moderator temperature coefficient, and differential control rod worth is shown to be significant for two PWRs of considerably different design

  2. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    Boldyrev, V.M.; Mikhaj, V.I.

    1985-01-01

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  3. Study on neutronics performance of flower shape advanced supercritical water cooled fast reactor with different solid moderators

    Yu Tao; Li Zhifeng; Xie Jinsen; Peng Honghua

    2015-01-01

    The supercritical water cooled fast reactors worked at such harsh condition with high temperature and high pressure, huge hydrogen balance pressure and thermal shock can result in a great loss of hydrogen. The released hydrogen would be out of control under accident situations. K_e_f_f, conversion ratio, moderator temperature effect, Doppler effect and void effect of different material such as ZrH_1_._7, Bp, BeO, C and SiC are discussed. BeO and SiC hold better integrated performance among these materials. Besides, moderators have less effect on the Doppler effect of fuel. (authors)

  4. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    Barariu, G.; Giumanca, R.

    2006-01-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the

  5. Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

    Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

    2002-01-01

    A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

  6. Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

    Bretscher, M.M.; Matos, J.E.

    1996-01-01

    At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U 3 Si 2 fuel is about 6.0 g U/cm 3 . The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm 3 and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm 3 . On the other hand, UN is the least reactive fuel because of the relatively large 14 N(n,p) cross section. For a fixed value of k eff , the required 235 U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO 2 dispersions are only useful for uranium densities below 5.0 g/cm 3 . In this density range, however, UO 2 is more reactive than U 3 Si 2

  7. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  8. Study of the heavy water regeneration processes

    Cavcic, E.

    1965-11-01

    Experience derived from heavy water reactor operation showed degradation and dilution of heavy water to be inevitable and depends on the type of reactor. Dilution of heavy water during operation of the RA and the RB reactors is shown in this report. Principles and procedures of heavy water regeneration by electrolysis, fractional distillation, cleaning, prevention of tritium contamination are described as well as separation columns

  9. Finishing and upgrading of heavy water

    Butler, J.P.; Hammerli, M.

    1981-01-01

    This invention provides a process and apparatus for deuterium enrichment as a final stage in a heavy water plant, for continuous on-line enrichment of the heavy water in moderator and heat transfer systems in heavy water nuclear reactors, and for enrichment of hevy water that has been downgraded with natural water during the course of operating a heavy water nuclear reactor. The method comprises contacting partially-enriched heavy water feed in a catalyst column with hydrogen gas (essentially D 2 ) orginating in an electrolysis cell so as to enrich the feed water with deuterium extracted from the electrolytic hydrogen gas and passing the deuterium-enriched water to the electrolysis cell. The apparatus comprises a catalyst isotope exchange column with hydrogen gas and liquid water passing through in countercurrent isotope exchange, an electrolysis cell, a dehumidifer-scrubber; and means for passing the liquid water enriched in deuterium from the catalyst column through the dehumidifer-scrubber to the electrolysis cell, for passing the hydrogen gas evolved in the cathode side of the cell through the dehumidifier-scrubber to the catalyst column, for passing the hydrogen gas from the catalyst column to an output, for introducing an input water feed to the upper portion of the catalyst column, and for taking a product enriched in deuterium from the system. (LL)

  10. Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

    Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

    1984-01-01

    The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

  11. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

    2005-08-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

  12. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Kukreja, Mukesh

    2005-01-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  13. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  14. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  15. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  16. Ultrasonic evaluation of end cap weld joints of fuel elements of pressurized heavy water reactors using signal analysis methods

    Raj, B.; Thavasimuthu, M.; Subramanian, C.V.; Kalyanasundaram, P.; Rajagopalan, C.

    1992-01-01

    This paper describes the application of ultrasonic digital signal analysis for the detection of fine defects of the order of 10% or lower of wall thickness (WT) of 370 microns in the resistance welded end cap-cladding tube joints of fuel elements used in Pressurised Heavy Water Reactors (PHWR s). The results obtained for the detection of such defects, have confirmed the sensitivity and reliability of this approach, and were further validated by destructive metallography. (author)

  17. Accident sequences evaluation using SFATs for low power and shutdown operation of pressurized heavy water reactors

    Kim, Chansoo; Chung, Chang-Hyun; Yang, Huichang

    2004-01-01

    To maintain the level of defense-in-depth safety of Pressurized Heavy Water Reactor (PHWR) during LP/SD operation, the qualitative risk evaluation methods such as Safety Function Assessment Trees (SFATs) are required. Therefore SFATs are suggested to assess and manage the PHWR safety in LP/SD. Before this study, safety functions of PHWR were classified into 7 groups; Reactivity Control, Core Cooling, Secondary Heat Removal, Primary Heat Transport Inventory, Essential Electrical Power, Cooling Water, and Containment Integrity. The SFATs for PHWR LP/SD operations were developed along with the Plant Outage Status (POS) variation, and totally 38 SFATs were developed for Wolsung Unit 2. For the verification of SFATs logics developed, top 5 accident sequences those contribute the CDF of PHWR were selected, and plant safety status were evaluated for those accident sequences. Accident sequences such as DCC-4 (Dual Control Computer Failure), CL4-16 (Total Loss of Class IV Power), and FWPV-11 (Loss of Feedwater Supply to SG due to Failure of Pumps/Values) were included. In this research the evaluation of plant safety status by accident sequences using SFATs and the verification of the SFATs were performed. Through the verification of SFAT logics, the enhancements to the internal logics of the SFATs were made, and the dependencies between safety systems and support systems were considered. It is expected the defense-in-depth evaluation model of PHW just as SFATs can be utilized in the configuration risk management program (CRMP) development and improve technical specifications development for Korean PHWRs. (author)

  18. Water cooled reactor technology: Safety research abstracts no. 1

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  19. An investigation of differences between measured and calculated bucklings of a series of light water and heavy water moderated experimental cores

    Figgins, A.J.G.

    1966-02-01

    A series of light water and light and heavy water moderated exponential and critical experiments performed by the Babcock and Wilcox Company were analysed using the METHUSELAH programme and it was found that the calculated and measured critical bucklings differed significantly. The effect was most marked as the temperature of the moderator was raised in the light water cores where it amounted to 10 m -2 for a 200 deg. C rise above room temperature. Of this discrepancy 3 m -2 , at the most, could be explained as being caused by the experimental cores not being large enough to have a central asymptotic region, leaving an unexplained difference of 7 m -2 . It is suggested that the only region in which METHUSELAH could be usefully modified to improve this agreement is in the calculation of the resonance escape probability. The last section of the report compares the calculated and measured results obtained at room temperatures. (author)

  20. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  1. Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

    Sienicki, J.J.; Horak, W.C.

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

  2. Study of the light and heavy water leaks in nuclear reactors and development of techniques for their detection, location and estimation

    Bashiruddin Mahmood, S.

    1979-01-01

    In heavy water type nuclear reactors the detection and control of heavy water and light water escapes from different systems is of vital importance in the successful and economic operation of these type of plants. The high cost of heavy water makes it imperative to minimise all such escapes, in order to reduce the loss as well as the upgrading cost of downgraded collection recovered from the reactor building. Original methods and devices have been developed at the Karachi Nuclear Power Plant which successfully solve this problem. This report describes the constructional and operational features of these devices

  3. Studies of the Effect of Heavy Water in the Fast Reactor FR0

    Tiren, L I; Haakansson, R; Karmhag, B

    1968-08-15

    Core 9 of the FR0 fast critical assembly was diluted with heavy water to 24 vol. per cent, contained in thin walled copper cans. The report describes measurements of the critical mass and the reactivity coefficient of heavy water in this core. The effect of the heterogeneous core composition on these items is also dealt with. The results are compared with theoretical predictions using several computer codes. Criticality is accurately predicted, but the measured reactivity coefficient of heavy water is about 20 % lower than the value obtained with the best available methods, involving the SPENG and DTF-4 programmes. The result of bunching measurements, in which the degree of heterogeneity of core composition was changed, is compared with theoretical estimates of the resonance shielding, flux advantage and leakage components of the heterogeneity effect.

  4. Modeling the transport of hydrogen in the primary coolant of pressurized heavy water reactors

    Subramanian, H.; Velmurugan, S.; Narasimhan, S.V.; Jain, A.K.; Dash, S.C.

    2008-01-01

    Heavy water (D 2 O) is used in primary heat transport systems of PHWRs. To suppress the radiolysis of heavy water and to control oxygen, hydrogen is added at regular intervals to the primary heat transport system. The added hydrogen finds it way to the heavy water storage tank after passing through the bleed condenser. Owing to the different temperatures and two phase region present in these systems, hydrogen gets redistributed. It is important to know the concentration of dissolved hydrogen in these regions in order to ensure a steady state dissolved hydrogen concentration in the primary system. Different power stations report variations in the frequency and quantity of hydrogen added to achieve the prescribed steady state level. This paper makes an attempt to account for the inventory of hydrogen and model its transport in PHT system. (author)

  5. Studies of the Effect of Heavy Water in the Fast Reactor FR0

    Tiren, L.I.; Haakansson, R.; Karmhag, B.

    1968-08-01

    Core 9 of the FR0 fast critical assembly was diluted with heavy water to 24 vol. per cent, contained in thin walled copper cans. The report describes measurements of the critical mass and the reactivity coefficient of heavy water in this core. The effect of the heterogeneous core composition on these items is also dealt with. The results are compared with theoretical predictions using several computer codes. Criticality is accurately predicted, but the measured reactivity coefficient of heavy water is about 20 % lower than the value obtained with the best available methods, involving the SPENG and DTF-4 programmes. The result of bunching measurements, in which the degree of heterogeneity of core composition was changed, is compared with theoretical estimates of the resonance shielding, flux advantage and leakage components of the heterogeneity effect

  6. Inspection of the UO2 special fuel for the prototype heavy water reactor 'FUGEN'

    Miura, Makoto; Ohmori, Takuro; Yoshino, Hiroyuki; Matsui, Hiromasa; Hirosawa, Naonori

    1979-01-01

    UO 2 special fuel assemblies are the fuel for material irradiation incorporating irradiation specimens, for the prototype heavy water reactor ''FUGEN''. In order to monitor the behavior of the pressure tube material irradiated with neutrons for long time, monitoring specimens were equipped in the core. This special fuel was fabricated by the Nuclear Fuel Industries, Ltd. (NFI), and the fuel cladding tubes, the capsule guide tubes and the capsule tubes were furnished by PNC. The irradiation specimens were prepared by PNC, and incorporated into the assemblies by NFI. The inspection by PNC on the special fuel assemblies was conducted following the inspection by the maker, which was made on UO 2 pellets, fuel element and assembly parts except cladding tubes, after completing the fabrication. The specifications of the special fuel, especially for the outer and inner layer pellets, the outer and inner layer fuel elements and the fuel assemblies, are presented. The flow sheet for the inspection process and surveillance test of special fuel assemblies is illustrated. The inspection items, the materials and the quantity inspection are tabulated for the fuel elements, the fuel assemblies and the irradiation capsules, respectively. The structure of a special type fuel assembly is shown. For each inspection, the inspection methods and items and the results are explained. As for the results of inspection of the special fuel, the UO 2 pellets, fuel element parts, fuel elements, fuel assembly parts, fuel assemblies, capsules and irradiation specimens were in accordance with the specifications. Regarding the situation of the quality control in the processes, check was made with many documents, and it was recognized that the quality control was performed in the quality assurance program. (Nakai, Y.)

  7. Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11; Prilog 11- Merenje nivoa teske vode u gorivnim kanalima reaktora RA

    Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was {+-}1 cm.

  8. Performance of refractometry in quantitative estimation of isotopic concentration of heavy water in nuclear reactor

    Dhole, K.; Roy, M.; Ghosh, S.; Datta, A.; Tripathy, M.K.; Bose, H.

    2013-01-01

    Highlights: ► Rapid analysis of heavy water samples, with precise temperature control. ► Entire composition range covered. ► Both variations in mole and wt.% of D 2 O in the heavy water sample studied. ► Standard error of calibration and prediction were estimated. - Abstract: The method of refractometry has been investigated for the quantitative estimation of isotopic concentration of heavy water (D 2 O) in a simulated water sample. Feasibility of refractometry as an excellent analytical technique for rapid and non-invasive determination of D 2 O concentration in water samples has been amply demonstrated. Temperature of the samples has been precisely controlled to eliminate the effect of temperature fluctuation on refractive index measurement. The method is found to exhibit a reasonable analytical response to its calibration performance over the purity range of 0–100% D 2 O. An accuracy of below ±1% in the measurement of isotopic purity of heavy water for the entire range could be achieved

  9. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  10. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  11. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  12. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  13. Secondary cycle water chemistry for 500 MWe pressurised heavy water reactor (PHWR) plant: a case study

    Bhandakkar, A.; Subbarao, A.; Agarwal, N.K.

    1995-01-01

    In turbine and secondary cycle system of 500 MWe PHWR, chemistry of steam and water is controlled in secondary cycle for prevention of corrosion in steam generators (SGs), feedwater system and steam system, scale and deposit formation on heat transfer surfaces and carry-over of solids by steam and deposition on steam turbine blades. Water chemistry of secondary side of SGs and turbine cycle is discussed. (author). 8 refs., 2 tabs., 1 fig

  14. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  15. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

    2003-01-01

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  16. Analytical performance of refractometry in quantitative estimation of isotopic concentration of heavy water in nuclear reactor

    Dhole, K.; Ghosh, S.; Datta, A.; Tripathy, M.K.; Bose, H.; Roy, M.; Tyagi, A.K.

    2011-01-01

    The method of refractometry has been investigated for the quantitative estimation of isotopic concentration of D 2 O (heavy water) in a simulated water sample. Viability of Refractometry as an excellent analytical technique for rapid and non-invasive determination of D 2 O concentration in water samples has been demonstrated. Temperature of the samples was precisely controlled to eliminate effect of temperature fluctuation on refractive index measurement. Calibration performance by this technique exhibited reasonable analytical response over a wide range (1-100%) of D 2 O concentration. (author)

  17. Preliminary definition of the design of a nuclear reactor for research and radioisotope production using natural uranium and heavy water

    Llagostera Beltran, J.I.

    1982-01-01

    A study was conducted about the evolution of the Brazilian importations of radioisotopes, from the beginning of the 70's since they have been increasingly used in the Country. In view of the limited production capacity of radioactive isotopes now existing in Brazil, a nuclear reactor type (natural uranium and heavy water) was defined, for research and production of radioisotopes, wich, besides providing, at least partially, the Brazilian needs of said isotopes, permits a large national participation in its project, construction and operating maintenance. The processes for heavy water production have been analyzed and it could be detected what is the best alternative for the production thereof, in low scale, in Brazil. The options concerning the definition of the main components of the reactor were justified and its most important features were determined, in relation to the neutronic and thermal aspects, being so defined its most significant parameters. The annual quantities were estimated, in terms of total and specific activity, for the radioisotopes that could be obtained by means of the proposed reactor, which, by now, are participating, to a large extent, in the total of Brazilian importation of radioactive isotopes. (Author) [pt

  18. Reactivity and reaction rate ratio changes with moderator voidage in a light water high converter reactor lattice

    Chawla, R.; Gmuer, K.; Hager, H.; Seiler, R.

    1986-01-01

    Integral reaction rate ratios and other k ∞ related parameters have been measured in the first three cores of the experimental program on light water high converter reactor (LWHCR) test lattices in the PROTEUS reactor. The reference tight-pitch lattice consisted of two rod types, with an average fissile-plutonium enrichment of 6% and a fuel/moderator ratio of 2.0. The moderators were H 2 O, Dowtherm (simulating an H 2 O voidage of 42.5%), and air (100% void). Comparisons of the measured parameters have been made with calculational results based mainly on the use of two separate codes and their associated data libraries, namely, WIMS-D and EPRI-CPM. A reconstruction of individual components of the k-infinity void coefficient has been carried out on the basis of the measured changes with voidage of the various reaction rate ratios, as well as of k-infinity itself. The subsequent more detailed comparisons between experiment and calculation should provide a useful basis for resolving the conflicting calculational results that have been reported in the past for the void coefficient characteristics of LWHCRs. (author)

  19. Development of in-situ laser based cutting technique for shock absorber rear nut in pressurized heavy water reactors. CP-2.1

    Vishwakarma, S.C.; Jain, R.K.; Upadhyaya, B.N.; Choubey, Ambar; Agrawal, D.K.; Oak, S.M.

    2007-01-01

    We have developed a laser based cutting technique for shock absorber rear nuts in pressurized heavy water reactors (PHWRs). This technique has been successfully used for in-situ laser cutting at RAPS-3 reactor. The technique consists of a motorized compact fixture, which holds a fiber optic beam delivery cutting nozzle and can be operated remotely

  20. Economic and safety aspects of using moderator heat for feed water heating in a nuclear power plant

    Patwegar, I.A.; Dutta, Anu; Chaki, S.K.; Venkat Raj, V.

    2002-01-01

    Full text: In the proposed advanced heavy water reactor (AHWR), coolant and moderator are separated by the coolant channel. The coolant absorbs most of the fission heat produced in the reactor core. However, the moderator absorbs about 5 to 6 % of the fission heat. In a reactor producing 750 MW(th) power, this moderator heat is about 40 MW. In the present Indian PHWR (pressurized heavy water reactor) systems, this moderator heat is lost to a sink through the moderator heat exchangers, which are cooled by process water. This paper presents the results of the steam cycle analysis carried out for AHWR using moderator heat exchangers as part of the feed heating system. The present study is an attempt to determine the gain in electrical output (MW) if moderator heat is utilized for feed water heating. The operational and safety aspects of using moderator heat are also discussed in the paper

  1. Tritium concentration in the heavy water upgrading plants

    Croitoru, C.; Pop, F.; Titescu, Gh.; Dumitrescu, M.; Ciortea, C.; Stefanescu, I.; Peculea, M.; Pitigoi, Gh.; Trancota, D. . E-mail of corresponding author: croitoru@icsi.ro; Croitoru, C.)

    2005-01-01

    In the course of time heavy water used in CANDU nuclear power plants, as moderator or coolant, degrades, as a result of its impurification with light water and tritium. Concentration diminution below 99.8% mol for moderator and 99.75% mol for coolant causes an inefficient functioning of CANDU reactor. By isotopic distillation, light water is removed. Simultaneously tritium concentration takes place. The heavy water upgrading plant from Cernavoda is an isotopic separation cascade with two stages. The paper presents, for this plant, a theoretical study of the tritium concentration. (author)

  2. Measurement of concentration of heavy water

    Tsukamoto, Yuichi; Kondo, Mitsuo; Sakurai, Naoyuki

    1979-01-01

    The concentration of heavy water is measured as one of the technical management in the Fugen plant. The heavy water is used as the moderator in the reactor. The measuring method depends on the theory of light absorption. The light absorption range of heavy water spreads from near infrared to infrared zone. The near infrared absorption was adopted for the purpose, as the absorption is much larger in infrared zone, and the measurement has to be conducted, limiting the apparent absorption. This measuring method is available to determine the concentration of heavy water in the broad range exactly. The preparation of heavy water sample and the measurement of the absorption spectra of near infrared ray are explained, as the experimental procedure. The sample cell was made of quartz, and the spectroscope was the Hitachi 323 type. The resolving power is 100 nm and 27 nm for the wave length of 1000 nm and 2500 nm, respectively. Concerning the measured results, the absorption was recorded in the wave length range from 600 nm to 2600 nm, and for the heavy water concentration range from 0 to 99.77 wt. %. The peaks of absorption were located at the wave length of 1450, 1660, 1920, 1970, 2020 and 2600 nm. The three kinds of fundamental vibration mode of the molecules of both light and heavy water are shown, and the peaks belong to H 2 O, HDO and D 2 O, respectively. The relation between the absorption and the heavy water concentration, and that between the transmissivity and the wave length are shown, when the cell thickness was varied to 5 mm and 20 mm, and the heavy water concentration to 21%, 62% and 99.85%. (Nakai, Y.)

  3. Peak power and heavy water production from electrolytic H2 and O2 using CANDU reactors

    Hammerli, M.; Stevens, W.H.; Bradley, W.J.; Butler, J.P.

    1976-04-01

    A combined energy storage - heavy water production system is presented. Off-peak nuclear energy is stored in the form of electrolytic H 2 (and O 2 ) from which a large fraction of the deuterium has been transferred to water in an H 2 /H 2 O deuterium exchange catalytic column. The main features and advantages of the combined electrolysis -catalytic exchange D 2 O process are discussed. Significant quantities of D 2 O could be produced economically at reasonable peak to base power cost ratios. Thirty to forty percent of the primary electric energy should be available for peak energy via either gas-steam turbines or fuel cells. (author)

  4. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  5. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  6. Numerical analysis and optimisation of heavy water upgrading column

    Sankar, Rama; Ghosh, Brindaban; Bhanja, K.

    2013-01-01

    In the 'Pressurised Heavy Water' type of reactors, heavy water is used both as moderator and coolant. During operation of reactor downgraded heavy water is generated that needs to be upgraded for reuse in the reactor. When the isotopic purity of heavy water becomes less than 99.75%, it is termed as downgraded heavy water. Downgraded heavy water also contains impurity such as corrosion products, dirt, oil etc. Upgradation of downgraded heavy water is normally done in two steps: (i) Purification: In this step downgraded heavy water is first purified to remove corrosion products, dirt, oil, etc. and (ii) Upgradation of heavy water to increase its isotopic purity, this step is carried out by vacuum distillation of downgraded heavy water after purification. This project is aimed at mathematical modelling and numerical simulation of heavy water upgrading column. Modelling and simulation studies of the upgradation column are based on equilibrium stage model to evaluate the effect of feed location, pressure, feed composition, reflux ratio in the packed column for given reboiler and condenser duty of distillation column. State to stage modelling of two-phase two-component flow has constitutes the overall modelling of the column. The governing equations consist of stage-wise species and overall mass continuity and stage-wise energy balance. This results in tridigonal matrix equation for stage liquid fractions for heavy and light water. The stage-wise liquid flow rates and temperatures are governed by stage-wise mass and energy balance. The combined form of the corresponding governing equations, with the incorporation of thermodynamic equation of states, form a system of nonlinear equations. This system have been resolved numerically using modified Newton-Raphson method. A code in the MATLAB platform has been developed by on above numerical procedure. The optimisation of the column operating conditions is to be carried out based on parametric studies and analysis of different

  7. Detailed description of an SSAC at the facility level for light water moderated (off-load refueled) power reactor facilities

    Jones, R.J.

    1985-03-01

    This report is intended to provide the technical details of an effective State Systems of Accounting for and Control of Nuclear Material (SSAC) which Member States may use, if they wish, to establish and maintain their SSACs. It is expected that systems designed along the lines described would be effective in meeting the objectives of both national and international systems for nuclear material accounting and control. This document accordingly provides a detailed description of a system for the accounting for and control of nuclear material in an off-load refueled light water moderated power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  8. Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

    Vishwakarma, S.C.; Upadhyaya, B.N.

    2016-01-01

    We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

  9. Possibilities of using metal uranium fuel in heavy water reactors; Mogucnosti upotrebe metalnog urana kao goriva za teskovodne reaktore

    Djuric, B; Mihajlovic, A; Drobnjak, Dj [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions. Postoje ozbiljni ekonomski razlozi za primenu metalnog urana u teskovodnim reaktorima, pre svega zbog njegove velike gustine, odnosno visokog konverzionog faktora, i zbog niskih troskova proizvodnje gorivnih elemenata. Glavne prepreke su bubrenje pri velikim stepenima sagorevanja i opasnost od korozije. Postoje veliki izgledi da se primenom odredjenih projektnih koncepcija i upotrebom legiranjem poboljsanog urana postigne zadovoljavajuca stabilnost metalnog urana u uslovima rada reaktora (author)

  10. RA Reactor operation and maintenance (I-IX), part V, Task 3.08/04-06, Refurbishment of the heavy water pumps

    Zecevic, V.; Nikolic, M.; Milic, J.

    1963-12-01

    In addition to detailed instructions for maintenance and repair of the heavy water pumps at the RA reactor this document includes nine annexes. They are as follows: cleaning the heavy water pump Avala with distilled water; instructions for repair of the pump CEN-132 (two annexes); list of operating characteristics of the pumps before repair; conclusions of the experts concerning the worn out bearings of the heavy water pump Avala, with the analysis of the stellite layer; report on the completed repair actions on the pumps Avala and CEN-132; report on the measurements done on the pump Avala; and the certificate concerning inspection of the pump

  11. Molten fuel-coolant interactions resulting from power transients in aluminium plate/water moderated reactors

    Storr, G.J.

    1989-08-01

    The behaviour of two reactors SL1 and SPERT D12, which underwent fast nuclear power transients prior to core destruction by a molten fuel-coolant interaction (MFCI) has been analysed and the results compared with measured data. The calculated spatial melt distribution and the mechanical work done during the events leads to high (∼ 250 kJ/kg) conversion efficiencies for this type of interaction when compared with molten drop experiments. A simple model for the steam explosion, using static thermodynamic properties of high temperature and pressure steam is used to calculate the dynamics of the reactors following the MFCI. 26 refs., 5 figs., 5 tabs

  12. Development of an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated reactors

    Babaev, N.S.

    1981-06-01

    The results of work carried out under IAEA Contract No. 2336/RB are described (subject: an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated (VVER) reactors). The basic principles of an accounting system for this type of nuclear power plant are outlined. The general structure and individual units of the information computer program used to achieve automated accounting are described and instructions are given on the use of the program. A detailed example of its application (on a simulated nuclear power plant) is examined

  13. Removal of gadolinium nitrate from heavy water

    Wilde, E.W.

    2000-03-22

    Work was conducted to develop a cost-effective process to purify 181 55-gallon drums containing spent heavy water moderator (D2O) contaminated with high concentrations of gadolinium nitrate, a chemical used as a neutron poison during former nuclear reactor operations at the Savannah River Site (SRS). These drums also contain low level radioactive contamination, including tritium, which complicates treatment options. Presently, the drums of degraded moderator are being stored on site. It was suggested that a process utilizing biological mechanisms could potentially lower the total cost of heavy water purification by allowing the use of smaller equipment with less product loss and a reduction in the quantity of secondary waste materials produced by the current baseline process (ion exchange).

  14. Problems of two-phase flows in water cooled and moderated reactors

    Syu, Yu.

    1984-01-01

    Heat exchange in two-phase flows of coolant in loss of coolant accidents in PWR and BWR reactors has been investigated. Three main stages of accident history are considered: blowdown, reflooding using emergency core cooling system and rewetting. Factors, determining the rate of coolant leakage and the rate of temperature increase in fuel cladding during blowdown, processes of vapour during reflooding and liquid priming by vapour during rewetting, are discussed

  15. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  16. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

  17. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-01-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  18. STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

    A. V. Korolyev

    2017-01-01

    list of initial events of severe accidents at NPPs with a water-cooled and water-moderated power reactor can be expanded.

  19. Safety aspects of long term operation of water moderated reactors. Recommendations on the scope and content of programmes for safe long term operation. Final report of the extrabudgetary programme on safety aspects long term operation of water moderated reactors

    2007-07-01

    During the last two decades, the number of IAEA Member States giving high priority to continuing the operation of nuclear power plants beyond the time frame originally anticipated is increasing. This is related to the age of nuclear power plants connected to the grid worldwide. The IAEA started to develop guidance on the safety aspects of ageing management in the 1990s. Recognizing the development in a number of its Member States, the IAEA initiated this Extrabudgetary Programme on Safety Aspects of Long Term Operation of Water Moderated Reactors in 2003. The objective of the Programme was to establish recommendations on the scope and content of activities to ensure safe long term operation of water moderated reactors. The term long term operation is used to accommodate various approaches in Member States and is defined as operation beyond an initial time frame set forth in design, standards, licence, and/or regulations, that is justified by safety assessment, considering life limiting processes and features for systems, structures and components. The scope of the Programme included general long term operation framework, mechanical components and materials, electrical components and instrumentation and control, and structural components and structures. The scope of the Programme was limited to physical structures of the NPPs. Four working groups addressed the above indicated technical areas. The Programme steering committee provided coordination and guidance and served as a forum for the exchange of information. The Programme implementation relied on voluntary in kind and financial contributions from the Czech Republic, Hungary, Slovakia, Sweden, the United Kingdom and the USA as well as in kind contributions from Bulgaria, Finland, the Netherlands, the Russian Federation, Spain, the Ukraine, and the European Commission. This report summarizes the main results, conclusions and recommendations of this Programme and provides in the Appendices I-IV detailed

  20. Super critical water reactors

    Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

    2005-01-01

    Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

  1. Trend of R and D publications in pressurised heavy water reactors: A study using INIS and other databases

    Kumar, V.; Kalyane, V.L.; Prakasan, E.R.; Kumar, A.; Sagar, A.; Mohan, L.

    2004-01-01

    Digital databases INIS (1970-2002), INSPEC (1969-2002), Chemical Abstracts (1977-2002), ISMEC (1973-June 2002), Web of Sciences (1974-2002), and Science Citation Index (1982-2002), were used for comprehensive retrieval of bibliographic details of research publications on Pressurized Heavy Water Reactor (PHWR) research. Among the countries contributing to PHWR research, India (having 1737 papers) is the forerunner followed by Canada (1492), Romania (508) and Argentina (334). Collaboration of Canadian researchers with researchers of other countries resulted in 75 publications. Among the most productive researchers in this field, the first 15 are from India. Top three contributors to PHWR publications with their respective authorship credits are: H.S. Kushwaha (106), Anil Kakodkar (100) and V. Venkat Raj (76). Prominent interdomainary interactions in PHWR subfields are: Specific nuclear reactors and associated plants with General studies of nuclear reactors (481), followed by Environmental sciences (185), and Materials science (154). Number of publications dealing with Geosciences aspect of environmental sciences are 141. Romania, Argentina, India and Republic of Korea have used mostly (≥75%) non-conventional media for publications. Out of the 4851 publications, 1228 have been published in 292 distinct journals. Top most journals publishing PHWR papers are: Radiation Protection and Environment (continued from: Bulletin of Radiation Protection since 1997), India (115); Nuclear Engineering International, UK (84); and Transactions of the American Nuclear Society, USA (68). (author)

  2. Improvement of lifetime availability through design, inspection, repair and replacement of coolant channels of Indian Pressurized Heavy Water Reactors

    Rupani, B.B.; Sinha, R.K.

    1998-01-01

    This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

  3. Activities in Argentina related to the use of slightly enriched uranium in heavy water reactor NPPs

    Corcuera, R.

    1999-01-01

    An overview of activities related to the use of Slightly Enriched Uranium (SEU) fuel in HWR type NPPs, currently under execution in Argentina, is presented. The activities here described cover certain R and D lines as well as the main aspects of the Project 'Transition from full Natural-U to full SEU core in Atucha-I NPP'. Concerning the R and D lines, a summary is given on investigations related to reduction of void-coefficient using SEU fuel assemblies, annular pellet SEU fuel for bundle power flattening, etc. The main aspects of the above mentioned Project are outlined. At present, Atucha-I core is approaching a 40% of core load with SEU fuel, while the target of full SEU fuel core should be reached in 2-3 years. The expected exit burnup for such a core, namely 11000 MWD/tnU, is already currently obtained for SEU fuel in the present mixed core, while an increase in exit burnup of Natural-U fuel has also been obtained in very good agreement with reactor physics calculations. The comprehensive safety analysis carried out for each phase of the Project showed very moderated changes in plant behaviour under the set of postulated accidents and abnormal transients. A recent development, namely the CARA Project, aimed at unifying manufacturing of fuel assemblies for both operating NPPs in Argentina is presented in an accompanying paper. (author)

  4. Application of fuzzy logic control system for reactor feed-water control

    Iijima, T.; Nakajima, Y.

    1994-01-01

    The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

  5. Fire damp gas in a heavy water reactor; Praskavi gas u teskovodnom reaktoru

    Nikolic, V D [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1963-07-01

    This document describes the process of fire damp gas creation in the reactor core and dependence of the gas percentage on the temperature, i.e. reactor power. It contains a detailed plan for measuring the the percent of fire damp gas at the RA reactor: before start-up, after longer shut-down periods, immediately after safety shutdown, periodically during operation campaign.

  6. Fast neutron flux in heavy water reactors; Flux de neutrons rapides dans les piles a eau lourde

    Brisbois, J; Katz, S [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

    1966-07-01

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author) [French] La possibilite de calculer le flux de neutrons rapides dans un reseau d'uranium naturel a eau lourde par superposition des apports des divers barreaux, a ete verifiee en utilisant un code Monte-Carlo monodimensionel. Les resultats obtenus concordent avec des mesures experimentales effectuees dans le coeur et reacteur de la pile Aquilon. (auteurs)

  7. Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

    Kim, Jong Sung; Oh, Young Jin

    2014-01-01

    If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

  8. Studies on the behaviour of a passive containment cooling system for the Indian advanced heavy water reactor

    Maheshwari, N.K.; Saha, D.; Chandraker, D.K.; Kakodkar, A.; Venkat Raj, V.

    2001-01-01

    A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser's working is the potential degradation of immersed condenser's performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to understand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V 1 and V 2 reached almost steady value. With varying energy input rate V 1 pressure dropped below the pressure in V 2 . The system could efficiently purge air from V 1 to V 2 . The paper deals with the details of the tests conducted and the results obtained. (orig.) [de

  9. Estimation of large early release frequency for Advanced Heavy Water Reactor

    Santhosh, T.V.; Thangamani, I.; Shrivastava, A.; Vinod, Gopika; Verma, Vishnu; Vijayan, P.K.

    2015-01-01

    Level 2 probabilistic safety assessment (PSA) examines severe accidents through a combination of probabilistic and deterministic approaches, in order to determine the release of radionuclides from containment, including the physical processes that are involved in the loss of structural integrity of the reactor core. The probabilistic part focuses on the reliability evaluation of containment systems and the deterministic part focuses on the analysis of the physical processes of an accident (timing and magnitude of radioactivity release), and the response of the containment. The important tasks involved are: (i) grouping and categorization of accident sequences into plant damage states (PDSs) (ii) development of a containment event trees (CETs) (iii) development of CET top event definitions and quantification of failure probabilities, and (iv) assigning the release categories and estimation of large early release frequency (LERF). LERF is defined as the frequency of those accidents leading to rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of offsite emergency response and protective actions such that there is the potential for early health effects. Such accidents generally include unscrubbed releases associated with early containment failure shortly after vessel breach, containment bypass events, and loss of containment isolation. Preliminary assessment of LERF, based on release categorization from qualitative expert judgement, has been carried out and the estimated LERF is found to be 1e-13/yr. The dominant contributors are: (a) LB LOCA with the failure of prompt shutdown coupled with containment isolation failure, (b) containment bypass event from main steam line break outside containment coupled with failure of main steam isolation valves, and (c) LB LOCA with complete failure of emergency core cooling system (ECCS) and loss of moderator cooling

  10. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  11. Calorimeter measurements of absorbed doses at the heavy water enriched uranium reactor

    Markovic, V.

    1961-12-01

    Application of calorimetry measurements of absorbed doses was imposed by the need of good knowledge of the absorbed dose values in the reactor experimental channels. Other methods are considered less reliable. The work was done in two phases: calorimetry measurements at lower reactor power (13-80 kW) by isothermal calorimeter, and differential calorimeter constructions for measurements at higher power levels (up to 1 MW). This report includes the following four annexes, papers: Isothermal calorimeter for reactor radiation monitoring, to be published; Calorimeter dosimetry of reactor radiation, presented at the Symposium about nuclear fuel held in april 1961; Radiation dosimetry of the reactor RA at Vinca, published in the Bull. Inst. Nucl. Sci. 1961; Differential calorimeter for reactor radiation dosimetry

  12. Conceptual design of a large heavy water reactor for US siting

    Shapiro, N.L.; Jesick, J.F.; Molin, A.T.; Daniel, J.A.

    1979-09-01

    Information on the PHWR type reactor is presented concerning design characteristics; fuel management and resource utilization; economic evaluations; safety, licensing, and environmental impact; and commercial introduction

  13. The security management of spent filter cartridge in Qinshan phase 3 (heavy water reactor) nuclear power plant

    Xue Dahai

    2005-01-01

    Qinshan phase 3 nuclear power plant is the first CANDU plant that China fetched in from Canada, and both two units operate under well condition up to now. The radioactive wastes produced during the unit operation mainly include technical waste, spent resin, and spent filter cartridge. The spent filter cartridge is one important part both in the volume and radioactivity of the radioactive waste, and it is the important content of radioactive waste management. Different from PWR, part of high radioactive spent filter in CANDU unit comes from heavy water system such as moderator system. It has to be dried through blowing before replaced from the system. But this working procedure result the filtrate dreg become flexible, and it can bring on the risk of internal or external exposure. It is very important to pay high attention to control the contamination spread during spent filter inside transfer. (authors)

  14. Optimal management of fuel in nuclear reactors with slightly enriched uranium and heavy water

    Serghiuta, D.

    1994-01-01

    This Ph.D. thesis presents the general principles guiding the optimal management of the fuel in CANDU type reactors with slightly enriched uranium. A method is devised which is based on the specific physical characteristics of this type of reactors and makes use of the multipurpose mathematical programming satisfying economical and nuclear safety requirements. The main goal of this work was the establishing of a refueling optimal methodology at equilibrium maintaining the reactor critical during operation. It also minimizes the fuel cycle cost through minimization of the utilized fissile material and at the same time by maximizing the reactor duty time through an optimal chain of refilling operations. This work can be considered as a contribution to a future project of CANDU type reactor core based on slightly enriched uranium. 74 Figs., 9 Tabs., 62 Refs

  15. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

    2009-03-01

    Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

  16. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-01-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  17. Production of heavy water

    Spencer, Larry S.; Brown, Sam W.; Phillips, Michael R.

    2017-06-06

    Disclosed are methods and apparatuses for producing heavy water. In one embodiment, a catalyst is treated with high purity air or a mixture of gaseous nitrogen and oxygen with gaseous deuterium all together flowing over the catalyst to produce the heavy water. In an alternate embodiment, the deuterium is combusted to form the heavy water. In an alternate embodiment, gaseous deuterium and gaseous oxygen is flowed into a fuel cell to produce the heavy water. In various embodiments, the deuterium may be produced by a thermal decomposition and distillation process that involves heating solid lithium deuteride to form liquid lithium deuteride and then extracting the gaseous deuterium from the liquid lithium deuteride.

  18. Power flattening and reactivity suppression strategies for the Canadian supercritical water reactor concept

    McDonald, M.; Colton, A.; Pencer, J.

    2015-01-01

    The Canadian supercritical water-cooled reactor (SCWR) is a conceptual heavy water moderated, supercritical light water cooled pressure tube reactor. In contrast to current heavy water power reactors, the Canadian SCWR will be a batch fuelled reactor. Associated with batch fuelling is a large beginning-of-cycle excess reactivity. Furthermore, radial power peaking arising as a consequence of batch refuelling must be mitigated in some way. In this paper, burnable neutron absorber (BNA) added to fuel and absorbing rods inserted into the core are considered for reactivity management and power flattening. A combination of approaches appears adequate to reduce the core radial power peaking, while also providing reactivity suppression. (author)

  19. Aspects on optimization of natural uranium fuel utilization in heavy water reactors

    1978-08-01

    This paper is dealing with a possibility to decrease the natural uranium consumption of CANDU PHWR using the once-through cycle. This possibility is based on the utilization of slightly enriched uranium. The optimal two-zone structure of a reactor using natural uranium is found out. The optimal criterium is the maximization of the burnup (equivalent to minimization of uranium requirements) with a constraint on power density radial uniformity factor. As regards the enriched uranium, the optimal enrichment and the two-zone structure of a reactor which minimizes the natural uranium requirement with constraints on uniformity factor and maximum burnup are established. Corresponding to a maximum burnup of 16,000 MWd/t and 1% enrichment, the natural uranium requirement is found to be 10% less than that of the natural uranium reactor

  20. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  1. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Lowry, N.J.

    1998-01-01

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints

  2. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  3. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-01-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  4. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  5. Heavy water: a distinctive and essential component of CANDU

    Miller, A.I.; van Alstyne, H.M.

    1994-06-01

    The exceptional properties of heavy water as a neutron moderator provide one of the distinctive features of CANDU reactors. Although most of the chemical and physical properties of deuterium and protium (mass 1 hydrogen) are appreciably different, the low terrestrial abundance of deuterium makes the separation of heavy water a relatively costly process, and so of considerable importance to the CANDU system. World heavy-water supplies are currently provided by the Girdler-Sulphide process or processes based on ammonia-hydrogen exchange. Due to cost and hazard considerations, new processes will be required for the production of heavy water in and beyond the next decade. Through AECL's development and refinement of wetproofed catalysts for the exchange of hydrogen isotopes between water and hydrogen, a family of new processes is expected to be deployed. Two monothermal processes, CECE (Combined Electrolysis and Catalytic Exchange, using water-to-hydrogen conversion by electrolysis) and CIRCE (Combined Industrially Reformed hydrogen and Catalytic Exchange, based on steam reforming of hydrocarbons), are furthest advanced. Besides its use for heavy-water production, the CECE process is a highly effective technology for heavy-water upgrading and for tritium separation from heavy (or light) water. (author). 10 refs., 1 tab., 7 figs

  6. Analysis of severe core damage accident progression for the heavy water reactor

    Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

    2010-01-01

    In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

  7. The possible use of cermet fuel in the DIDO and PLUTO heavy-water research reactors

    Kennedy, T.D.A.

    1981-08-01

    As part of a study of the feasibility of using low-enrichment fuels in DIDO and PLUTO reactors the heat transfer and safety aspects involved in replacing the present U/AL-alloy (75% w/w U 235 ) fuel plates with U/AL-cermet (20% w/w U 235 ) plates, having the same outside dimensions to retain the same hydraulic characteristics, have been investigated. (U.K.)

  8. Canadian heavy water production

    Dahlinger, A.; Lockerby, W.E.; Rae, H.K.

    1977-05-01

    The paper reviews Canadian experience in the production of heavy water, presents a long-term supply projection, relates this projection to the anticipated long-term electrical energy demand, and highlights principal areas for further improvement that form the bulk of our research and development program on heavy water processes

  9. Transient regimes in a heavy water reactor; Regimes transitoires dans un reacteur a eau lourde

    Raievski, V [Commissariat a l' Energie Atomique, Saclay(France). Centre d' Etudes Nucleaires

    1953-07-01

    We studied the variations of power and reactivity of a reactor when we raise in a continuous way the starting plates. During the subcritical regime (negative reactivity), the power is determined by reactivity and by the intensity of the sources of photo neutrons, produced during the previous work of the reactor. When, during the rise of the plates, the reactor, pass by the critical regime (zero reactivity), one notes that the reached power is independent of the initial reactivity. During the sur-critical regime (positive reactivity), the elevation of temperature of the uranium bars slows down the growth of reactivity due to the movements of the plates. The power stretches then toward a value that depends only on the regime of cooling of the reactor and the excess of the available reactivity. This survey permits to choose such a rise speed, that reactivity remains constantly lower to a value beyond which the piloting of the reactor becomes difficult. This result is not more valid, if the intensity of the sources is insufficient, what takes place during the first divergences and after a stop of long length. (author) [French] On etudie les variations de puissance et de reactivite d'un reacteur quand on leve d'une facon continue les plaques de demarrage. Pendant le regime subcritique (reactivite negative), la puissance est determinee par la reactivite et par l'intensite des sources de photoneutrons, produites pendant la marche anterieure du reacteur. Quand, au cours de la montee des plaques, le reacteur passe par le regime critique (reactivite nulle), on constate que la puissance atteinte est independante de la reactivite initiale. Pendant le regime surcritique (reactivite positive), l'elevation de temperature des barres d'uranium ralentit l'accroissement de reactivite due aux mouvements des plaques. La puissance tend alors vers une valeur qui ne depend plus que du regime de refroidissement du reacteur et de l'exces de la reactivite disponible. Cette etude permet de

  10. Detection of tritium in the air surrounding the heavy water reactors; Elementi detekcije tricijuma u vazduhu kod teskovodnih nuklearnih reaktora

    Ninkovic, M; Matic-Vukmirovic, Z; Hadzisehovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1967-03-15

    This paper contains the study of the literature concerned with physical properties of the tritium, problems of detection control of the tritium level in the atmosphere in the vicinity of heavy water reactors. It is stated that a complete and efficient control of tritium activity, from radiation protection point of view can be achieved only by simultaneous triple measurements: direct measurement of tritium in the air by stationary or movable instruments; air sampling and measurement of activity by laboratory instrumentation; and measurement of tritium in the bio-material of the personnel who have inhaled air contaminated with tritium. Laboratory equipment was adapted for tritium detection in air samples. A method for measuring the specific tritium activity was developed and implemented. The tritium level and distribution in the air were measured during exchange of the fuel channel in the RA reactor. The obtained results indicate that tritium could be dangerous for the staff involved. Proucena je literatura u kojoj se tretiraju osnovne fizicke karakteristike tricijuma, kao i problemi detekcije i kontrole u vazduhu kod teskovodnih nuklearnih reaktora. Utvrdjeno je da kompletna i efikasna kontrola aktivnosti tricijuma, sa aspekta zastite od zracenja, moze biti ostvarena samo ako se vrse istovremeno trostruka merenja: merenje aktivnosti tricijuma u vazduhu direktno, prenosnim ili stacioniranim instrumentima; uzimanje uzoraka vazduha i merenje aktivnosti na laboratorijskoj aparaturi; i merenje aktivnosti tricijuma u biomaterijalu osoblja koje je udisalo vazduh kontaminiran tricijumom. Izvrsena je adaptacija laboratorijske aparature za potrebe detekcije tricijuma u uzorcima vazduha. Razradjen je i uhodan postupak merenja koncentracije aktivnosti tricijuma u uzorcima vazduha. Izvrsena su merenja i dobijeni su rezultati o nivou i raspodeli tricijuma u vazduhu pri operaciji zamene kanala sa gorivom na reaktoru RA u Vinci. Dobijeni rezultati ukazuju na opasnost koju po radno

  11. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-01-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

  12. Heavy water lattices: Second panel report

    1963-09-15

    The panel was attended by prominent physicists from most of the laboratories engaged in the field of heavy water lattices throughout the world. The participants presented written contributions and status reports describing the past history and plans for further development of heavy-water reactors. Valuable discussions took place, during which recommendations for future work were formulated. Refs, figs, tabs.

  13. Heavy water lattices: Second panel report

    1963-01-01

    The panel was attended by prominent physicists from most of the laboratories engaged in the field of heavy water lattices throughout the world. The participants presented written contributions and status reports describing the past history and plans for further development of heavy-water reactors. Valuable discussions took place, during which recommendations for future work were formulated. Refs, figs, tabs

  14. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C.

    2017-01-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO_2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D_2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α"M_T(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  15. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  16. Behaviour of heavy water in nuclear reactors of the CEA; Comportement de l'eau lourde dans les piles du C.E.A

    Chenouard, J; Dirian, G; Roth, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    In the two heavy water reactors of the CEA: Zoe and P-2, we do: A) the supervision of the isotopic composition of the heavy water; B) the supervision of gases released by the decomposition of the heavy water under radiation, and to their recombination; C) periodic analyses of impurities. (M.B.) [French] Dans les deux piles a eau lourde du Commissariat a l'Energie Atomique: Zoe et P 2, nous effectuons: A) la surveillance de la composition isotopique de l'eau lourde; B) la surveillance des gaz degages par la decomposition de l'eau lourde sous radiation, et a leur recombinaison; C) des analyses periodiques d'impuretes. (M.B.)

  17. Chemical elimination of alumina in suspension in nuclear reactors heavy water; Elimination de l'alumine en suspension dans l'eau lourde des reacteurs nucleaires par voie chimique

    Ledoux, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-02-01

    Corrosion of aluminium in contact with moderating water in nuclear reactor leads to the formation of an alumina hydrosol which can have an adverse effect on the operation of the reactor. Several physical methods have been used in an attempt to counteract this effect. The method proposed here consists in the elimination of the aluminium by dissolution and subsequent fixation in the ionic form on mixed-bed ion-exchange resin. In order to do this, the parameters and the values of these parameters most favorable to the dissolution process have been determined. If the moderator is heavy water, the deuterated acid can be prepared by converting a solution in heavy water to a salt of the acid using a deuterated cationic resin. (author) [French] La corrosion de l'aluminium au contact de l'eau moderatrice des reacteurs nucleaires, donne lieu a la formation d'un hydrosol d'alumine nuisible au bon fonctionnement des reacteurs. Plusieurs methodes physiques ont ete mises en oeuvre pour pallier ces inconvenients. On propose ici d'eliminer l'alumine par solubilisation pour la fixer ensuite sous forme ionique par des resines echangeuses d'ions, en lit melange. A cette fin on determine les parametres et leurs grandeurs favorables a cette solubilisation. Si le moderateur est de l'eau lourde la preparation d'acide deutere peut etre effectuee par passage d'une solution en eau lourde a un sel de l'acide sur resine cationique deuteree.

  18. Pressure tube reactor

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

    Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

  19. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

    2010-01-01

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  20. Computed phase equilibria for burnable neutron absorbing materials for advanced pressurized heavy water reactors

    Corcoran, E.C. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada)], E-mail: emily.corcoran@rmc.ca; Lewis, B.J.; Thompson, W.T. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada); Hood, J. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada); Akbari, F.; He, Z. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ont., K0J 1J0 (Canada); Reid, P. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada)

    2009-03-31

    Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU [CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.

  1. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

  2. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  3. Determination of the tritium content in the reactor heavy water, Phase II; Odredjivanje porasta kolicine tritijuma u reaktorskoj teskoj vodi, II faza

    Ribnikar, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    Measurement results of the {sup 3}H activity in non-irradiated water and after reactor operation are presented. Methods were developed for sampling and radiochemical water purification by ion exchange and multiple distillation. Methods for absolute measurement of soft beta radiation of tritium were established. Migration of tritium through the heavy water RA reactor system was monitored. Results were compared with other measured reactor parameters. Prikazani su rezultati merenja aktivnosti {sup 3}H u nezracenoj vodi i posle rada reaktora; razradjeni su metodi za uzimanje i radiohemijsko preciscavanje vode putem jonske izmene i visestepene destilacije; postavljeni metodi za apsolutno merenje mekog beta-zracenja tritijuma; pracene su migracije tritijuma kroz teskovodni sistem reaktora; takodje su interpretirani i poredjeni rezultati sa drugim merenim parametrima reaktora.

  4. FEM analysis of foundation raft for 500 MWe pressurized heavy water reactor building

    Kulkarni, N.N.; Goray, J.S.; Joshi, M.H.; Paramasivam, V.

    1989-01-01

    Foundation raft supports the containment structure and internals for 500 MWe PHW reactor building. It also serves as bottom envelop of the containment structure. In view of this, the design of foundation raft assumes great importance. The foundation raft is subjected to various load, most significant of them are dead load of structure, equipment loads transferred through a system of floors, walls and structural steel columns, pressure load during accident conditions, seismic loads, earth pressure, uplift due to buoyancy loads, foundation reaction etc. In order to achieve optimum design, the detailed structural analysis is required to be performed methodically and in most realistic manner. Finite element methods which have come in vogue with the developments in digital computers can be successfully applied in this area. The paper describes the above methods in detail for the analysis of foundation raft for the various load combinations required to be considered for safe and optimum design

  5. Overview moderator material for nuclear reactor components

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

    In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

  6. Design of condenser for 500 MWe pressurised heavy water reactors (PHWRs) - a case study

    Agarwal, N.K.; Subbarao, A.; Chaudhary, K.

    1996-01-01

    Condenser forms the major heat sink in the power plants. In recent years, power plant availability and performance have become great concern to the industry. The detailed design of the condenser and its associated cooling water (CW) system require careful optimisation of parameters which include material selection, cooling water flow rate, condenser surface areas, turbine exhaust pressures etc. This is required to produce a design offering maximum efficiency and reliability and minimum maintenance. The various parameters involved in condenser design are discussed. 5 refs., 1 fig

  7. Intercomparison and validation of computer codes for thermalhydraulic safety analysis of heavy water reactors

    2004-08-01

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and co-operative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled: Intercomparison and validation of computer codes for thermalhydraulics safety analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. The RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommendations made

  8. Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

    2012-08-01

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

  9. Design and fabrication of fuel for the prototype heavy water reactor Fugen

    Hasumi, Takashi; Yamanaka, Ryozi; Osawa, Masahide; Asami, Tomohiro; Kaziyama, Takashi

    1983-01-01

    For the advanced thermal reactor Fugen, 224 fuel assemblies were charged as the initial charge fuel, of which 96 were uranium-plutonium mixed oxide fuel, and 128 were uranium dioxide fuel. Since the full scale operation was started in March, 1979, fuel exchange was carried out five times, and 240 fuel assemblies were taken out, but fuel breaking was never found, and the fuel for Fugen has shown good result. For 16 mixed oxide fuel assemblies for the third exchange and thereafter, the domestically produced plutonium extracted in the Tokai reprocessing plant has been used, and for 15 UO 2 fuel assemblies for the fifth exchange, the enriched uranium produced in the Ningyo Pass plant was used. These fuels are burning in the core without causing trouble. The course of the development of the fuel is described as follows: trial manufacture, evaluation test outside the core, heat transferring flow characteristic test, irradiation test, design of fuel elements and fuel assemblies, production of fuel and quality assurance, and results of production and use. (Kako, I.)

  10. Study of the heavy water regeneration processes; Studija procesa za regeneraciju teske vode

    Cavcic, E [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    Experience derived from heavy water reactor operation showed degradation and dilution of heavy water to be inevitable depends on the type of reactor. Dilution of heavy water during operation of the RA and the RB reactors is shown in this report. Principles and procedures of heavy water regeneration by electrolysis, fractional distillation, cleaning, prevention of tritium contamination are described as well as separation columns.

  11. Determination of heavy water in heavy water - light water mixtures

    Sanhueza M, A.

    1986-01-01

    A description about experimental methodology to determine isotopic composition of heavy water - light water mixtures is presented. The employed methods are Nuclear Magnetic Resonance Spectroscopy, for measuring heavy water concentrations from 0 to 100% with intervals of 10% approx., and mass Spectrometry, for measuring heavy water concentrations from 0.1 to 1% with intervals of 0.15% approx., by means of an indirect method of Dilution. (Author)

  12. Development of quantitative analytical procedures on two-phase flow in tight-lattice fuel bundles for reduced-moderation light-water reactors

    Ohnuki, A.; Kureta, M.; Takae, K.; Tamai, H.; Akimoto, H.; Yoshida, H.

    2004-01-01

    The research project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR is a light water reactor for which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles whose gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes. (authors)

  13. Development of a noise-based method for the determination of the moderator temperature coefficient of reactivity (MTC) in pressurized water reactors (PWRs)

    Demaziere, C.

    2002-01-01

    The Moderator Temperature Coefficient of reactivity (MTC) is an important safety parameter of Pressurized Water Reactors (PWRs). In most countries, the so-called at-power MTC has to be measured a few months before the reactor outage, in order to determine if the MTC will not become too negative. Usually, the at-power MTC is determined by inducing a change in the moderator temperature, which has to be compensated for by other means, such as a change in the boron concentration. An MTC measurement using the boron dilution method is analysed in this thesis. It is demonstrated that the uncertainty of such a measurement technique is so large, that the measured MTC could become more negative than what the Technical Specifications allow. Furthermore, this technique incurs a disturbance of the plant operation. For this reason, another technique relying on noise analysis was proposed a few years ago. In this technique, the MTC is inferred from the neutron noise measured inside the core and the moderator temperature noise measured at the core-exit, in the same or in a neighbouring fuel assembly. This technique does not require any perturbation of the reactor operation, but was nevertheless proven to underestimate the MTC by a factor of 2 to 5. In this thesis, it is shown, both theoretically and experimentally, that the reason of the MTC underestimation by noise analysis is the radially loosely coupled character of the moderator temperature noise throughout the core. A new MTC noise estimator, accounting for this radially non-homogeneous moderator temperature noise is proposed and demonstrated to give the correct MTC value. This new MTC noise estimator relies on the neutron noise measured in a single point of the reactor and the radially averaged moderator temperature noise measured inside the core. In the case of the Ringhals-2 PWR in Sweden, Gamma-Thermometers (GTs) offer such a possibility since in dynamic mode they measure the moderator temperature noise, whereas in static

  14. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  15. Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

    Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2013-10-15

    Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  16. ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

    VIKAS JAIN

    2013-10-01

    Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  17. Advanced light-water reactors

    Golay, M.W.; Todreas, N.E.

    1990-01-01

    Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

  18. Evaluation of endcap welds in thin walled fuel elements of pressurised heavy water reactor by ultrasonic testing

    Subramanian, C.V.; Thavasimuthu, M.; Kalyansundaram, P.; Bhattacharya, D.K.; Raj, Baldev

    1992-01-01

    In the pressurised heavy water reactor systems of India, the fuel is encapsulated in thin-walled tubes (0.342 mm) closed with endcaps by resistance welding. The integrity of these fuel elements should be such that no fission gas leakage takes place during reactor operation. The quality control of the endcap welds needed to satisfy this requirement includes helium leak test and destructive metallographic test (on sample basis). This paper discusses the feasibility study that has been carried out in the author's laboratory to develop an immersion ultrasonic test method for evaluating the integrity of the endcap weld region. Through holes of various sizes (0.15mm, 0.2mm, 0.4mm diameter and 0.185mm and 0.342mm deep) were machined by spark erosion machining at the weld joints to simulate defects of various sizes. Line focussed probe of 10 MHz frequency was used for the testing. It was possible to detect clearly all the machined holes. Based on the above standardised procedure, further testing was done on endcap welds which were rejected during fabrication on account of showing leak rate of 3 x 10 -6 std. c.c/sec. or more during helium leak test. Though it was possible to get echoes from the natural defects in the rejected tubes with echo amplitude of 70%, the signal was accompanied by the geometrical reflection (noise) giving an amplitude of 20% from the weld region, giving rise to the problem of resolving the defect indication from the geometric indications. Therefore, signal analysis approach was adopted. The signal obtained from the weld zone were subjected to various analysis procedures like a) autopower spectrum, b) total energy content and c) demodulated auto correlation function. It was possible by all the three methods to differentiate the defect signal from those due to weld geometry or due to noise. Subsequently, metallography was carried out to characterise the type of defects observed during the ultrasonic testing. (author). 4 figs

  19. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    2006-05-01

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  20. Feasibility study on the thorium fueled boiling water breeder reactor

    PetrusTakaki, N.

    2012-01-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  1. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

    Roche, R; Gaudez, J C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration, l'equilibrage des pression entre l'eau lourde et le gaz, le montage

  2. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

    Roche, R.; Gaudez, J.C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration

  3. Development of a purification system at Dhruva to treat oil contaminated and chemically impure heavy water

    Suttraway, S.K.; Mishra, V.; Bitla, S.V.; Ghosh, S.K.

    2006-01-01

    Dhruva, a 100 MW (thermal) Research reactor uses Heavy Water as moderator, reflector and coolant. Normally during plant operation, the Heavy water from the system gets removed during operational and maintenance activities and this collected heavy water gets degraded and contaminated in the process. The degraded heavy water meeting the chemical specification requirement of the up gradation plant is sent for up gradation. Part of the Heavy water collected is contaminated with various organic and inorganic impurities and therefore cannot be sent for IP up gradation as it does not meet the chemical specification of the up gradation plant. This contaminated Heavy water was being stored in SS drums. Over the years of Reactor operation reasonable amount of contaminated Heavy water got collected in the plant. This Heavy water collected from leakages, during routine maintenance, operational activities and fuelling operation had tritium activity and variety of contamination including oil, chlorides, turbidity due to which the specific conductivity was very high. It was decided to purify this Heavy water in house to bring it up to up gradation plant chemical specification requirement. There were number of challenges in formulating a scheme to purify this Heavy water. The scheme needed to be simple and compact in design which could be set up in the plant itself. It should not pose radiological hazards due to radioactive Heavy water during its purification and handling. The contaminated Heavy water collected in drums had varying chemistry and IP. The purification plant should be able to do batch processing so that the different IP and chemical quality of Heavy water stored in different drums are not mixed during purification. It should be capable of removing the oil, chlorides, turbidity and decrease the conductivity to acceptable limits of the Up gradation plant. A purification plant was developed and commissioned after detail laboratory studies and trials. This paper explains

  4. Methodologies and technologies for life assessment and management of coolant channels of Indian pressurised heavy water reactors

    Rupani, B.B.; Sinha, S.K.; Sinha, R.K.

    2002-01-01

    Zirconium alloy coolant channels are central to the design of Indian Pressurised Heavy Water Reactors (PHWRs) and form the individual pressure boundaries. These coolant channels consist of horizontal pressure tubes made of zirconium alloys, which are separated from cold calandria tubes using garter spring spacers. High temperature heavy water coolant flows through the pressure tube which supports the fuel bundles. A typical coolant channel in a PHWR is shown. These pressure tubes are subjected to several life limiting degradation mechanisms like creep and growth, hydrogen pick-up, reduction in fracture toughness and delayed hydride cracking phenomena because of their operation under high temperature, high stress and high fast neutron flux environment. Considering the early onset of these degradation mechanisms in Zircaloy-2 pressure tubes used in the early generation of Indian PHWRs, the life management of these coolant channels becomes a challenging task, involving multidisciplinary R and D efforts in areas like analytical modelling of degradation mechanisms, evolution of methodologies for assessment of fitness for service and, tools and techniques for remote on line monitoring of integrity, maintenance and replacement. The degradation mechanisms have been modelled and incorporated into specially developed computer codes, such as SCAPCA for irradiation induced creep and growth deformation modelling, HYCON for hydrogen pick-up modelling, BLIST for hydrogen diffusion, blister nucleation and growth modelling and CEAL for assessment of leak before break behaviour. These codes have been validated with respect to the results of in-service inspection and post irradiation examination. Development of analytical models actually paved the way for the evolution of more refined methodologies for assessing the safe residual life of coolant channel. Information gathered from various experiments simulating the degradation mechanisms, results of post-irradiation examination of the

  5. High purity heavy water production: need for total organic carbon determination in process water streams

    Ayushi; Kumar, Sangita D.; Reddy, A.V.R.; Vithal, G.K.

    2009-01-01

    In recent times, demand for high purity heavy water (99.98% pure) in industries and laboratories has grown by manifold. Its application started in nuclear industry with the design of CANDU reactor, which uses natural uranium as fuel. In this reactor the purest grade of heavy water is used as the moderator and the primary coolant. Diverse industrial applications like fibre optics, medicine, semiconductors etc. use high purity heavy water extensively to achieve better performance of the specific material. In all these applications there is a stringent requirement that the total organic carbon content (TOC) of high purity heavy water should be very low. This is because the presence of TOC can lead to adverse interactions in different applications. To minimize the TOC content in the final product there is a need to monitor and control the TOC content at each and every stage of heavy water production. Hence a simple, rapid and accurate method was developed for the determination of TOC content in process water samples. The paper summarizes the results obtained for the TOC content in the water samples collected from process streams of heavy water production plant. (author)

  6. Nuclear power plant life management processes: Guidelines and practices for heavy water reactors. Report prepared within the framework of the Technical Working Groups on Advanced Technologies for Heavy Water Reactors and on Life Management of Nuclear Power Plants

    2006-06-01

    The time is right to address nuclear power plant life management and ageing management issues in terms of processes and refurbishments for long term operation and license renewal aspects of heavy water reactors (HWRs) because some HWRs are close to the design life. In general, HWR nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. This involves the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. This TECDOC deals with organizational and managerial means to implement effective plan life management (PLiM) into existing plant in operating HWR NPPs. This TECDOC discusses the current trend of PLiM observed in NPPs to date and an overview of PLiM programmes and considerations. This includes key objectives of such programs, regulatory considerations, an overall integrated approach, organizational and technology infrastructure considerations, importance of effective plant data management and finally, human issues related to ageing and finally integration of PLiM with economic planning. Also general approach to HWR PLiM, including the key PLiM processes, life assessment for critical structures and components, conditions assessment of structures and components and obsolescence is mentioned. Technical aspects are described on component specific technology considerations for condition assessment, example of a proactive ageing management programme, and Ontario power generation experiences in appendices. Also country reports from Argentina, Canada, India, the Republic of Korea and Romania are attached in the annex to share practices and experiences to PLiM programme. This TECDOC is primarily addressed to both the management (decision makers) and technical staff (engineers and scientists) of NPP owners/operators and technical support

  7. Pulse radiolysis studies of liquid heavy water at temperatures up to 250 degrees C

    Stuart, C.R.; Ouellette, D.C.; Elliot, A.J.

    2002-09-01

    This report documents the rate constants and associated activation energies for the reactions of the primary radical species, e aq - , ·OD and ·D, which are formed during the radiolysis of heavy water within the temperature range 20 to 250 o C. These heavy-water data have been compared with the corresponding information for light water. These kinetic data form part of the database that is required to model the aqueous radiation chemistry that occurs within the core of the heavy water cooled and moderated CANDU reactor. (author)

  8. Pulse radiolysis studies of liquid heavy water at temperatures up to 250 degrees C

    Stuart, C.R.; Ouellette, D.C.; Elliot, A.J

    2002-09-01

    This report documents the rate constants and associated activation energies for the reactions of the primary radical species, e{sub aq}{sup -}, {center_dot}OD and {center_dot}D, which are formed during the radiolysis of heavy water within the temperature range 20 to 250 {sup o}C. These heavy-water data have been compared with the corresponding information for light water. These kinetic data form part of the database that is required to model the aqueous radiation chemistry that occurs within the core of the heavy water cooled and moderated CANDU reactor. (author)

  9. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  10. Moderator circulation in CANDU reactors

    Fath, H.E.S.; Hussein, M.A.

    1989-01-01

    A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

  11. Problems related with the power regulation of reactors by physico-chemical methods, and the behaviour of water and heavy water in nuclear reactors; Comportement de l'eau et de l'eau lourde dans les reacteurs nucleaires et problemes de la regulation de puissance par voie physico-chimique

    Dolle, L; Conan, D; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Experience of the CEA heavy water reactors and a systematic study of the radiolytic decomposition of water in the core of swimming-pool reactors are described. Setting up of reactivity control by physico-chemical methods. Reactivity control by homogeneous poisoning of the reactor A comparison of the evolution of xenon poisoning with the residual anti reactivity of the poison in solution during its nuclear consumption establishes the programme which must govern the variation in its concentration if the exact compensation is to be produced The behaviour of the poison towards the reactor materials under the particular operational conditions must be taken into account. Radiolytic decomposition of water in the reactors in the presence of soluble poisons: A study of the effect of certain chemically inert salts, present in small concentrations in the water, on its radiolytic decomposition rate, has led to some new results which are discussed. The choice of a soluble poison is justified on the basis of the above results. Reactivity control by the use of a gaseous absorbent The use of a gas control rod circuit for compensation purposes, in place of solid control rods is described. The use of soluble poisons in the moderator to compensate the xenon effect, and of a gaseous absorbent in a circuit known as a gas control rod form original aspects of the reactivity control in the reactor EL 4. (authors) [French] L'observation du comportement de l'eau et de l'eau lourde dans les reacteurs en exploitation, contribue au fonctionnement sur de ceux-ci et oriente certaines etudes relatives aux techniques de controle de la reactivite par mise en oeuvre de poisons solubles. L'utilisation de poisons nucleaires dissous dans l'eau du reacteur entraine une pollution chimique de celle-ci. Les conditions d'emploi permettant d'eviter les effets indesirables de cette pollution sont etudiees. Les problemes analytiques - bien qu'importants - ne sont pas abordes dans le cadre de la communication

  12. Heavy water at Trail, British Columbia

    Arsenault, J.E. [Ontario (Canada)

    2006-09-15

    Today Canada stands on the threshold of a nuclear renaissance, based on the CANDU reactor family, which depends on heavy water as a moderator and for cooling. Canada has a long history with heavy water, with commercial interests beginning in 1934, a mere two years after its discovery. At one time Canada was the world's largest producer of heavy water. The Second World War stimulated interest in this rather rare substance, such that the worlds largest supply (185 kg) ended up in Canada in 1942 to support nuclear research work at the Montreal Laboratories of the National Research Council. A year later commercial production began at Trail, British Columbia, to support work that later became known as the P-9 project, associated with the Manhattan Project. The Trail plant produced heavy water from 1943 until 1956, when it was shut down. During the war years the project was so secret that Lesslie Thomson, Special Liaison Officer reporting on nuclear matters to C.D. Howe, Minister of Munitions and Supply, was discouraged from visiting Trail operations. Thomson never did visit the Trail facility during the war. In 2005 the remaining large, tall concrete exchange tower was demolished at a cost of about $2.4 million, about the same as it cost to construct the facility about 60 years ago. Thus no physical evidence remains of this historic facility and another important artifact from Canada's nuclear history has disappeared forever. It is planned to place a plaque at the site at some point in the future. (author)

  13. Heavy water at Trail, British Columbia

    Arsenault, J.E.

    2006-01-01

    Today Canada stands on the threshold of a nuclear renaissance, based on the CANDU reactor family, which depends on heavy water as a moderator and for cooling. Canada has a long history with heavy water, with commercial interests beginning in 1934, a mere two years after its discovery. At one time Canada was the world's largest producer of heavy water. The Second World War stimulated interest in this rather rare substance, such that the worlds largest supply (185 kg) ended up in Canada in 1942 to support nuclear research work at the Montreal Laboratories of the National Research Council. A year later commercial production began at Trail, British Columbia, to support work that later became known as the P-9 project, associated with the Manhattan Project. The Trail plant produced heavy water from 1943 until 1956, when it was shut down. During the war years the project was so secret that Lesslie Thomson, Special Liaison Officer reporting on nuclear matters to C.D. Howe, Minister of Munitions and Supply, was discouraged from visiting Trail operations. Thomson never did visit the Trail facility during the war. In 2005 the remaining large, tall concrete exchange tower was demolished at a cost of about $2.4 million, about the same as it cost to construct the facility about 60 years ago. Thus no physical evidence remains of this historic facility and another important artifact from Canada's nuclear history has disappeared forever. It is planned to place a plaque at the site at some point in the future. (author)

  14. Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

    2007-06-01

    be used in reassessing the safety of individual operating plants. In 1998, the IAEA completed IAEA-TECDOC-1044 entitled Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for their Resolution and established the associated LWRGSIDB database (Computer Manual Series No. 13). The present compilation, which is based on broad international experience, is an extension of this work to cover pressurized heavy water reactors (PHWRs). As in the case of LWRs, it is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It addresses generic safety issues identified in nuclear power plants using PHWRs. In most cases, the measures taken or planned to resolve these issues are also identified. The work on this report was initiated by the Senior Regulators of Countries Operating CANDU-Type Nuclear Power Plants at one of their annual meetings. It was carried out within the framework of the IAEA's programme on National Regulatory Infrastructure for Nuclear Installation Safety and serves to enhance regulatory effectiveness through the exchange of safety related information

  15. Safety system in a heavy water detritiation plant

    Balteanu, O.; Stefan, I.; Retevoi, C.

    2003-01-01

    In a CANDU 6 type reactor a quantity of 55·10 15 Bq/year of tritium is generated, 95% being in the D 2 O moderator which can achieve a radioactivity of 2.5-3.5·10 12 Bq/kg. Tritium in heavy water contributes with 30-50% to the doses received by operation personnel and up to 20% to the radioactivity released in the environment. The large quantity of heavy water used in this type of reactors (500 tones) make storage very difficult, especially for environment. The extraction of tritium from tritiated heavy water of CANDU reactors solve the following problems: the radiation level in the operation area, the costs of maintenance and repair reduction due to reduction of personnel protection measures, the increase of NPP utilisation factor by shutdown time reduction for maintenance and repair, use the extracted tritium for fusion reactors and not for the last, lower costs and risk for storage heavy water waste. Heavy water detritiation methods, which currently are used in the industrial or experimental plant, are based on catalytic isotope exchange or electrolysis followed cryogenic distillation or permeation. The technology developed at Institute of Cryogenics and Isotope Separation is based upon catalytic exchange between tritiated water and deuterium, followed by cryogenic distillation of hydrogen isotopes. The nature of the fluids that are processed in detritiation requires the operation of the plant in safety conditions. The paper presents the safety system solution chose in order to solve this task, as well as a simulation of an incident and safety system response. The application software is using LabView platform that is specialised on control and factory automation applications. (author)

  16. Fundamentals of pressurized water reactors

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  17. Moderator heat recovery of CANDU reactors

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  18. An analysis of workers' tritium concentration in urine samples as a function of time after intake at Korean pressurised heavy water reactors.

    Kim, Hee Geun; Kong, Tae Young

    2012-12-01

    In general, internal exposure from tritium at pressurised heavy water reactors (PHWRs) accounts for ∼20-40 % of the total radiation dose. Tritium usually reaches the equilibrium concentration after a few hours inside the body and is then excreted from the body with an effective half-life in the order of 10 d. In this study, tritium metabolism was reviewed using its excretion rate in urine samples of workers at Korean PHWRs. The tritium concentration in workers' urine samples was also measured as a function of time after intake. On the basis of the monitoring results, changes in the tritium concentration inside the body were then analysed.

  19. Travelling cranes for heavy reactor component handling

    Champeil, M.

    1977-01-01

    Structure and operating machinery of two travelling cranes (600 t and 450 t) used in the Framatome factory for handling heavy reactor components are described. When coupled, these cranes can lift loads up to 1000 t [fr

  20. IR analyzer spots heavy water leaks

    Anon.

    1978-01-01

    A correlation spectrometer developed by Barringer Research Ltd. (in collaboration with Atomic Energy of Canada and Ontario Hydro) is used to measure HDO concentrations in DTO in the final (distillation) stage of heavy-water production. A unit has been installed at Bruce Heavy Water Plant. Previously, such spectrometers had been installed to detect heavy-water leaks in CANDU reactors. The principle on which the instrument works is explained, with illustrations. It works by comparing the absorption at 2.9 μm, due to HDO, with that at 2.6 μm, due to both HDO and D 2 O. (N.D.H.)

  1. Heavy water at Aswan

    1959-01-01

    A fertilizer factory is being built by Egyptian Chemical Industries (Kima) at Aswan on the upper Nile; it will produce a mixture of ammonium nitrate and calcium carbonate adjusted to contain 20.5% nitrogen. It is also proposed to construct a heavy water plant to be located at and integrated with the fertilizer factory. At the request of the Government of the United Arab Republic, the International Atomic Energy Agency sent an expert to carry out investigation of the technical, economic and other related aspects of the proposed production of heavy water. A report was submitted to the IAEA Director General. Its main conclusions can be summarized as follows: (1) Production of heavy water as a by-product of fertilizer manufacture at Aswan is technically feasible. Separation of deuterium from industrial hydrogen for this purpose could be done either by catalytic exchange or by liquefaction and distillation; the choice should depend on economic considerations. (2) The heavy water produced at Aswan should be competitive in cost with that produced elsewhere; this, however, would depend on whether firm contracts are obtained for the delivery of equipment at guaranteed prices and with guaranteed performance, and whether such prices are in reasonable agreement with preliminary estimates. (3) The future market for heavy water is difficult to predict. For one thing, there is a very large production capacity in the USA, most of which is idle due to lack of demand. Secondly, there is a relatively small production outside the USA that is sold at prices higher than that charged by the US Government. The future of the market is necessarily contingent upon the possibility of future free sale by the US Government. At the end of his report, the expert has also given his comments on possible further assistance to the project by IAEA

  2. Studies on gadolinium precipitation in moderator system of nuclear reactor

    Joshi, Akhilesh C.; Rajesh, Puspalata; Rufus, A.L.; Velmurugan, S.

    2015-01-01

    Gadolinium is used in the moderator system of many Pressurised Heavy Water Reactors (PHWRs) for start-up, shut-down and reactivity control during operation. It is very much essential to maintain gadolinium concentration in the system as desired. It has been reported that gadolinium gets precipitated in as oxalate in carbonated water under the influence of γ-radiation. Hence, studies were carried out to investigate the effect of dose, presence of other metal ions and metal surfaces on the precipitation of gadolinium. The results showed that the amount of carboxylic acids viz., formic acid and oxalic acid, formed due to radiolysis is dependent on the dose and that the curve passes though a maxima. Gadolinium is added in higher concentration in Advanced Heavy Water Reactor. So, experiments with high concentration of gadolinium were also carried out. Ultra pure water saturated with high purity CO 2 containing gadolinium and desired ion/surface was irradiated with γ-radiation from 60 Co source at 25°C to doses ranging from 2.5-16.6 Mrad. At lower doses, formation of carboxylic acids takes place but as the dose increases, decomposition of these acids starts and hence the concentration Vs dose passes through a maximum. It was found that precipitation of gadolinium as oxalate occurred at lower doses. At higher doses, it was seen that pH of the solution decreases and hence solubility of gadolinium oxalate increases. It was also observed that the amount of gadolinium precipitated varied linearly with the initial concentration of gadolinium varying from 2 ppm to 20 ppm. While for gadolinium concentration from 20 ppm to 400 ppm, gadolinium in particulate form was observed. The amount of carboxylic acids formed depends on the nature of cations present in solution. It was found that the amount of oxalic acid formed in the case of gadolinium was more than that formed in the case of sodium. Presence of metal oxides such as ZrO 2 formed over zircoloy surfaces was found to

  3. Historical summary of the heavy-section steel technology program and some related activities in light-water reactor pressure vessel safety research

    Whitman, G.D.

    1986-03-01

    The accomplishments of the Heavy-Section Steel Technology Program and other programs having a close relationship to the development of information used in the assessment of light-water reactor pressure vessel integrity are reviewed. The early Pressure Vessel Research Committee planning, the principals contributing to program formulation, the role of the US Atomic Energy Commission, and the developments under the US Nuclear Regulatory Commission sponsorship are identified. The need for major research and development accomplishments in fracture mechanics, heavy-section steel procurement, materials properties, irradiation effects, fatigue crack growth, and structural testing are summarized. The impact of program results on regulatory issues and the development of data used in the preparation of codes, standards, and guides are discussed. Continuing activities and recommendations for future research and development in support of pressure vessel integrity assessments are presented

  4. Light water reactor safety

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  5. A study on the application of CRUDTRAN code in primary systems of domestic pressurized heavy-water reactors for prediction of radiation source term

    Song, Jong Soon; Cho, Hoon Jo; Jung, Min Young; Lee, Sang Heon [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

  6. Containment for Heavy-Water Gas-Cooled Reactors; Le Confinement des Reacteurs a Eau Lourde Refroidis par Gaz

    Verstraete, P.; Lehmann, D.; Lafitte, R. [Bonard et Gardel, Ingenieurs-Conseils, Lausanne (Switzerland)

    1967-09-15

    The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author) [French] Les principes de securite des reacteurs a eau lourde refroidis

  7. Canadian heavy water production

    Dahlinger, A.; Lockerby, W.E.; Rae, H.K.

    1977-01-01

    The paper reviews Canadian experience in the production of heavy water, presents a long-term supply projection, relates this projection to the anticipated long-term electrical energy demand, and highlights principal areas for further improvement that form the bulk of the Canadian R and D programme on heavy water processes. Six Canadian heavy water plants with a total design capacity of 4000Mg/a are in operation or under construction. All use the Girdler-Sulphide (GS) process, which is based on deuterium exchange between water and hydrogen sulphide. Early operating problems have been overcome and the plants have demonstrated annual capacity factors in excess of 70%, with short-term production rates equal to design rates. Areas for further improvement are: to increase production rates by optimizing the control of foaming to give both higher sieve tray efficiency and higher flow rates, to reduce the incapacity due to deposition of pyrite (FeS 2 ) and sulphur (between 5% and 10%), and to improve process control and optimization of operating conditions by the application of mathematical simulations of the detailed deuterium profile throughout each plant. Other processes being studied, which look potentially attractive are the hydrogen-water exchange and the hydrogen-amine exchange. Even if they become successful competitors to the GS process, the latter is likely to remain the dominant production method for the next 10-20 years. This programme, when related to the long-term electricity demand, indicates that heavy water supply and demand are in reasonable balance and that the Candu programme will not be inhibited because of shortages of this commodity. (author)

  8. Heavy water leak detection using diffusion sampler

    Joshi, M.L.; Hussain, S.A.

    1990-01-01

    In the Pressurrised Heavy Water Reactors (PHWRs) detection of the sources of heavy water leaks is importent both for the purpose of radiation hazard control as well as for the reduction of escape/loss of heavy water which, is an expensive nuclear material. This paper describes an application of tritium diffusion sampler for heavy water leak detection. The diffusion sampler comprises an usual tritium counting glass vial with a special orifice. The counting vial has water vapour, deficient in HTO concentration. The HTO present outside diffuses in the vial through the orifice, gets exchanged with water of the wet filter paper kept at the bottom and the moisture in the vial atmosphere which has HTO concentration lower than that outside. This results in continuation of net movement of HTO in the vial. The exchanged tritium is counted in liquid scintillation spectrometer. The method has a sensitivity of 10000 dpm/DAC-h. (author). 2 figs., 2 ta bs

  9. Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

    Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

  10. The high moderating ratio reactor using 100% MOX reloads

    Barbrault, P.

    1994-06-01

    This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

  11. From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca; Od kriticnog sistema teska voda - prirodni uranium do brzo - termickog istrazivackog reaktora u Vinci

    Stefanovic, D; Pesic, M [Vinca Institute of Nuclear Sciences, Beograd (Yugoslavia)

    1995-07-01

    A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

  12. Neutrinos: Heavy water detector

    Anon.

    1990-01-01

    The proponents of the Sudbury Neutrino Observatory (SNO) received a welcome Christmas present when William Winegard, Canadian Minister for Science and Technology announced the final details of the funding for this project, totalling 48 million Canadian dollars and including contributions from the US and the UK. The SNO experiment will extend significantly the study of solar neutrinos, using some 1,000 tonnes of heavy water to be installed more than two kilometres below ground in a nickel mine at Sudbury, Ontario

  13. Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

    Sapra, M.K.; Kundu, S.; Pal, A.K.; Sharma, B.S.V.G.

    2007-02-01

    During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

  14. Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

  15. The importance of heavy water in nuclear technology

    Gharib, A.G.

    2004-01-01

    Due to similarities of chemical and almost physical properties in H 2 O and D 2 O but differences in nuclear and particle peculiarities provide valuable application for D 2 O. To sustain a controlled chain reaction, the energy of neutrons produced by fission must be reduced through collisions with other nuclei, a process called moderation. A good moderator has a mass close to that of the neutron to maximize energy loss per collision and a very small neutron capture cross section to minimize unwanted nuclear reactions. Deuterium is far the best moderator, more than 80 times better than hydrogen and 30 times better than 12 C ir 18 O. Heavy water is almost as good as deuterium and has the distinct advantage of being a nonflammable liquid. Heavy water is also an excellent neutron reflector, and thus decreases the number of neutrons that escape the reactor core without participating in fission reactions. For this reason a feasibility study and subsequently a technical survey was carried out on engineering of a pilot scale plant. As the result of this studies, the know-how of heavy water production on basis of selected method including dual temperature isotopic exchange and distillation techniques developed. Subsequently the primary and almost detail engineering documents prepared on best knowledge of our own engineers without external contribution

  16. Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

    Gupta, S.K.; Gaikwad, A.J.; Kumar, Rajesh

    2004-01-01

    From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

  17. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    Yilmaz, S.

    2000-01-01

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  18. Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2016-11-15

    Highlights: • Simplified models with inclusion of the clad temperature are considered. • Boiling nonlinearity and core power removal have been modeled. • Method of multiple time scales has been used for nonlinear analysis to get the nature and amplitude of oscillations. • Incorporation of modeling complexities enhances the stability of system. • We find that reactors with higher nominal power are more desirable from the point of view of global stability. - Abstract: We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all

  19. Nuclear reactor types

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  20. A feasibility study on long-life reduced-moderation water reactor with highly protected Pu breeding by doping with minor actinides

    Hamase, Erina; Damian, Frederic; Poinot-salanon, Christine; Saito, Masaki; Sagara, Hiroshi; Han, Chi Young

    2013-01-01

    Highlights: ► The present paper is focused on a reduced-moderation water reactor. ►237 Np or 241 Am was doped into the inner blanket. ►238 Pu transmuted from MA works as a fissionable nuclide. ► It also contributes to increasing the proliferation resistance of Pu. ► Long-life core and highly protected Pu breeding were simultaneously achieved. - Abstract: The present paper is focused on the feasibility of reduced-moderation water reactor (RMWR) which could simultaneously achieve the extension of core life-time, and highly protected plutonium (Pu) breeding performance with short compound system doubling time (CSDT) by neptunium-237 ( 237 Np) or americium-241 ( 241 Am) doping of the inner blanket of the RMWR. As a preliminary analysis of the RMWR, a simplified fuel pin configuration was analyzed. In case of 60% doping of 237 Np and 241 Am in the inner blanket, the maximum available effective full power days (EFPDs) were much longer and were about 15,100 and 14,200 EFPDs, respectively. For the proliferation resistance of Pu produced in the blanket, the total Pu in the inner and lower/upper blanket was below the practically unusable criterion for an explosive device proposed by Pellaud, and was technically unfeasible for high-technology hypothetical nuclear explosive devices (HNEDs) proposed by Kessler and Kimura. The protected Pu breeding performance was analyzed and in case of 60% doping of 237 Np and 241 Am, CSDT was short by approximately 40 and 50 years. Therefore, the feasibility of RMWR was shown which could simultaneously achieve the extension of core life-time and have highly protected Pu breeding performance with short CSDT by doping the inner blanket with minor actinides (MAs). The objective of this study was thus to evaluate the performance of the concept with respect to the core life-time, proliferation resistance of Pu and CSDT. The technological feasibility of such core concept will have to be evaluated by further dedicated analyses

  1. The Bare Critical Assembly of Natural Uranium and Heavy Water

    Popovic, D [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1958-07-01

    The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

  2. Supercritical Water Reactors

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  3. Pressurised water reactor operation

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  4. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries; Definicion y Analisis de Benchmarks de Reactores de Agua Pesada para Pruebas de Nuevas Bibliotecas de Datos Wims-D

    Leszczynski, Francisco [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

    2000-07-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable.

  5. Assessment of competing reaction effect on results of activation analysis with use of water-cooled and water-moderated reactor neutron fields

    Avsaragov, Kh.B.; Toichkin, A.N.; Lobov, A.N.

    1988-01-01

    Effect of competing threshold reactions on results of neutron activation analysis (NAA) using WWER-440 reactor is investigated. (n,p) and (n,α) fast neutron and 232 Th (n,f), 235 U(n,f), 238 U(n,f) fast and thermal neutron processes are considered as competing ones. Contribution of competing reactions when determining Na, Mn, Sc, Fe, Cu, Y for the core channels of in-core monitoring and ionization chamber ring water protection is experimentally evaluated using a spectrometer with Ge(Li) detector in a set with AI-4096 analyser. Under rigid neutron fields interfering activity increases at the expense of thorium and uranium atom fission. It is stressed that when determining Zr, Mo, Ru, Ba, La, Ce, Nd contribution of fission reaction products can appear to be sufficient

  6. Study of the modifications on the synchronous generators, heavy water pumps and condenser batteries of the RA reactor - Annex 17; Prilog 17 - Elaborat o izmenama u semama sinhronih generatora, teskovodnih pumpi i kondenzatorskih baterija reaktora RA

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    Modifications done on the synchronous generators are related to the emergency power supply system, meaning one of the most important devices responsible for reactor safety. Without reducing the efficiency of the heavy water pumps the improved stability of generators operation was achieved by reducing the possibility of errors and simplifying manipulation. Condensator batteries were improved in order to decrease the leakage currents.

  7. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

  8. Fissile fuel assembly for a sub-moderated nuclear reactor

    Millot, J.P.; Dejeux, Pol.; Alibran, Patrice.

    1983-01-01

    Each of the core assemblies is composed of a prismatic case made of a neutron absorbing material, inside which very long rods containing the fissile material are arranged parallel to the height of the case and according to a regular network in the straight sections of the case. At least one piece in a fertile material exposed to the neutrons emitted by the fissile material of the assembly is arranged on each one of the side faces of the case. The invention applies in particular to sub-moderated reactors, cooled and moderated by pressurized water [fr

  9. Reactor water sampling device

    Sakamaki, Kazuo.

    1992-01-01

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  10. Electrolytic process for upgrading heavy water (Preprint No. PD-16)

    Rammohan, K.; Sadhukhan, H.K.

    1989-04-01

    In the reactor system the heavy water gets depleted in concentration due to leakages, intermixing and vapour collection in boiler vault system etc. Electrolysis of water was used as a secondary plant to enrich the dilute heavy water produced in the primery plant by hydrogen-sulfide-water exchange process. The studies made in the development of this process for the upgrading of downgra ded heavy water by setting up a full size Electrolyser Test Assembly are discussed a nd complete design of a heavy water upgrading plant based on electrolytic process for MAPS and NAPP is described. (author). 7 refs., 5 figs

  11. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

  12. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  13. Fluid moderator control system reactor internals distribution system

    Fensterer, H.F.; Klassen, W.E.; Veronesi, L.; Boyle, D.E.; Salton, R.B.

    1987-01-01

    This patent describes a spectral shift pressurized water nuclear reactor employing a low neutron moderating fluid for the spectral shift including a reactor pressure vessel, a core comprising a plurality of fuel assemblies, a core support plate, apparatus comprising means for penetrating the reactor vessel for introducing the moderating fluid into the reactor vessel. Means associated with the core support plate for directly distributing the moderating fluid to and from the fuel assemblies comprises at least one inlet flow channel in the core plate; branch inlet feed lines connect to the inlet flow channel in the core plate; vertical inlet flow lines flow connected to the branch inlet feed lines; each vertical flow line communicates with a fuel assembly; the distribution means further comprise lines serving as return flow lines, each of which is connected to one of the fuel assemblies; branch exit flow lines in the core plate flow connected to the return flow lines of the fuel assembly; and at least one outlet flow channel flow connected to the branch exit flow lines; and a flow port interposed between the penetration means and the distribution means for flow connecting the penetration means with the distribution means

  14. Reactors of different types in the world nuclear power

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  15. Status and perspective of development of cold moderators at the IBR-2 reactor

    Kulikov, S; Shabalin, E

    2012-01-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  16. Status and perspective of development of cold moderators at the IBR-2 reactor

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  17. Nuclear reactors. Introduction

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  18. Neutronic study of the two french heavy water reactors; Etude neutronique des deux piles francaises a eau lourde

    Horowitz, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [French] Les deux reacteurs francais - la pile de Chatillon, appelee ZOE, et la pile de Saclay, designee dans la suite par P2 - ont fait l'objet d'etudes neutroniques detaillees dont les principales sont exposees dans ce rapport. Ces etudes ont ete pour la plupart effectuees dans le cadre du Departement des Etudes de Piles (D.E.P.). Nous avons ainsi entre autre etudie la distribution du flux neutronique; les facteurs influencants la reactivite; le lien entre reactivite et divergence par la formule de Nordheim; le temps de vie moyen des neutrons; les spectres de neutrons de P2; l'effet xenon; ou encore l'effet des differents reglages des plaques et barres de controles. (M.B.)

  19. Heavy Water - Industrial Separation Processes

    Peculea, M.

    1984-01-01

    This monograph devoted to the heavy water production mainly presents the Romanian experience in the field which started in early sixties from the laboratory scale production and reached now the level of large scale industrial production at ROMAG-Drobeta, Romania. The book is structured in eleven chapters entitled: Overview, The main physical properties, Sources, Uses, Separation factor and equilibrium constant, Mathematical modelling of the separation process, Thermodynamical considerations on the isotope separation, Selection criteria for heavy water separation processes, Industrial installations for heavy water production, Prospects, Acknowledgements. 200 Figs., 90 Tabs., 135 Refs

  20. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  1. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

    2012-07-01

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

  2. Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2; Prilog 2 - Misljenje o teskocama eksploatacije reaktora RA u uslovima visoke aktivnosti teskovodnog sistema

    Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events. [Serbo-Croat] Zakljuceno je da je eksploatacija reakora u uslovima postojece aktivnosti teskovodnog sistema veoma opasna po sam reaktor i po personal. Dva osnovna razloga u prilog ovog zakljucka su: nedovoljna mogucnost kontrole ispravnosti svih elemenata reaktora u toku remonta i ekspolatacije u predstojecem periodu; teskoca borbe sa udesima, koji bi se eventualno dogodili podjednako je opasna po instalaciju i okolinu. Pre no sto se nastavi sa daljom ekspolatacijom reaktora potrebno ciscenje, dekontaminacija sistema teske vode kako bi se izbegla moguca ostecenja ili akcidentalne situacije.

  3. European supercritical water cooled reactor

    Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

    2011-01-01

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  4. Reactor water chemistry control

    Kundu, A.K.

    2010-01-01

    Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

  5. The Canadian heavy water situation

    Dahlinger, A.

    The Canadian heavy water industry is analyzed. Supply and demand are predicted through 1985. Pricing is broken down into components. Backup R and D contributes greatly to process improvements. (E.C.B.)

  6. A study of Cirus heavy water system isotopic purity

    Thomas, Shibu; Sahu, A.K.; Unni, V.K.P.; Pant, R.C.

    2000-01-01

    Cirus uses heavy water as moderator and helium as cover gas. Approximately one tonne of heavy water was added to the system every year for routine make up. Isotopic purity (IP) of this water used for addition was always higher than that of the system. Though this should increase IP of heavy water in the system, it has remained almost at the same level, over the years. A study was carried out to estimate the extent of improvement in IP of heavy water in the system that should have occurred because of this and other factors in last 30 years. Reasons for non-occurrence of such an improvement were explored. Ion exchange resins used for purification of heavy water and air ingress into helium cover gas system appear to be the principal sources of entry of light water into heavy water system. (author)

  7. The Canadian heavy water situation

    Dahlinger, A.

    1975-08-01

    Existing heavy water plants in Canada are producing at a satisfactory rate and currently planned capacity is in balance with projected needs. By 1980, we shall have Girdler-Sulphide plants installed with a design capacity of almost 600 kg/h. Effort is required to minimize production costs for heavy water and to ensure that costs do not increase faster than the current inflationary trend. (Author)

  8. Verification of HELIOS-MASTER system through benchmark of Halden boiling water reactor (HBWR)

    Kim, Ha Yong; Song, Jae Seung; Cho, Jin Young; Kim, Kang Seok; Lee, Chung Chan; Zee, Sung Quun

    2004-01-01

    To verify the HELIOS-MASTER computer code system for a nuclear design, we have been performed benchmark calculations for various reactor cores. The Halden reactor is a boiling, heavy water moderated reactor. At a full power of 18-20MWt, the moderator temperature is 240 .deg. C and the pressure is 33 bar. This study describes the verification of the HELIOS-MASTER computer code system for a nuclear design and the analysis of a hexagonal and D 2 O moderated core through a benchmark of the Halden reactor core. HELIOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves the neutronics model with the TPEN (Triangle based Polynomial Expansion Nodal) method for a hexagonal geometry

  9. Water cooled nuclear reactor

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  10. Reactor core

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  11. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  12. Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

    Liem, Peng Hong; Sembiring, Tagor Malem

    2012-01-01

    Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

  13. Boiling water reactor fuel bundle

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

  14. Transmutation of Tc-99 in fission reactors

    Kloosterman, J.L.; Li, J.M.

    1994-12-01

    Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

  15. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2011-07-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  16. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

    2011-01-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  17. WATER BOILER REACTOR

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  18. The pressurized water reactor

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  19. Reactor water level control device

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  20. Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

    Permana, S.; Takaki, N.; Sekimoto, H.

    2007-01-01

    Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

  1. Fuel to Moderator Ratio Sensitivity Study Using Water Rod Moderator in SCWR Conceptual Core Design

    Bae, Seong Man; Kim, Yong Bae; Park, Dong Hwan; Lee, Kwang Ho

    2009-01-01

    The conceptual operating condition of Super Critical Water-cooled Reactor (SCWR) is above critical point of water, such that the coolant temperature ranges from 280 .deg. C to 510 .deg. C with a pressure of 25MPa. This operating condition makes an SCWR have both merits and demerits when compared with current Light Water Reactors (LWRs). One of the demerits of SCWR is degradation of neutron moderation due to a lower water density from ∼0.1g/cm 3 to ∼0.7g/cm 3 under a high coolant temperature condition. Therefore it is necessary to enhance the moderation capability for SCWR to slow down the fast fission neutrons. Many SCWR designs have a water rod concept as an additional moderator, because water has a good moderation capability. Through reviewing the previous water rod assembly designs, it was identified that a sensitivity study is required to optimize fuel assembly pitch to increase the neutron economy. In this paper, the results of the conceptual assembly design sensitivity study which focuses on the comparison of sensitivity for the fuel pitch to diameter (P/D) ratio using water rod moderator, are presented

  2. A progress review of Ontario Hydro's nuclear generation and heavy water production programs

    Kee, F.J.; Woodhead, L.W.

    Performance and economics of CANDU reactors in service are described. Progress of commissioning, construction and planning of reactors at Pickering, Bruce, and Darlington is outlined. Heavy water production is reviewed. (E.C.B.)

  3. Heavy water plant

    Rogers, D.G.

    1978-01-01

    This invention provides an auxiliary contactor column or exchange tower to receive stripped gas and vapour from a stripper. An auxiliary supply of heated feed water is passed in isotope exchanging relation with the gas in the auxiliary contactor to raise the deuterium content of the gas, which then is returned to the main process, at the hot tower or at the feed absorption tower as already described in relation to previous practice. Flow balance between gas and water in the auxiliary contactor is achieved relatively simply by monitoring the deuterium content of the hot water leaving the contactor column, and regulating the supply of hot water, to the contactor column in response thereto. (author)

  4. Pressurized-water reactors

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  5. Advanced boiling water reactor

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  6. A simple and rapid gas chromatographic method for the determination of dissolved deuterium and nitrogen in heavy water coolant of a nuclear reactor

    Nair, B.K.S.

    1976-01-01

    A known volume of a heavy water sample is equilibrated with a known volume of pure helium gas at atmospheric pressure in a sample tube. The dissolved gases evolve into the helium and distribute themselves between the gaseous and liquid phases according to their equilibrium partial pressures. These partial pressures of the gases in the equilibrium gas mixture are determined by analysing it gas-chromatographically. From these analytical data and the absorption coefficients of deuterium and nitrogen, their original concentrations in heavy water are calculated. Corrections for the increase in the total pressure of the gaseous phase owing to evolved gases are calculated and found to be negligible. Air contamination during sampling and analysis can be detected by the presence of the oxygen peak in the chromatogram and corrected for. The calculation is facilitated by programming it on an electronic calculator. The method is much simpler and faster than the vacuum method usually applied for this analysis. One determination can be completed in about an hour. The average deviation and standard deviation have been estimated at 0.19 ml/litre heavy water and 0.25 ml/litre heavy water respectively in deuterium, and 0.36 and 0.68 ml/litre in nitrogen. (author)

  7. Reactor water injection facility

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  8. Status of advanced technology and design for water cooled reactors: Light water reactors

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  9. Estimation of fracture toughness of Zr 2.5% Nb pressure tube of Pressurised Heavy Water Reactor using cyclic ball indentation technique

    Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.; Rama Rao, A.

    2016-08-15

    Highlights: • Measurement of fracture toughness of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situ Property Measurement System (IProMS) has been designed in house. • Conventional and IProMS tests conducted on pressure tube spool pieces having different mechanical properties. • Correlation has been established between the conventional and IProMS estimated fracture properties. - Abstract: In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes made up of Zr 2.5 wt% Nb alloy. Pressure tubes undergo degradation during its service life due to high pressure, high temperature and radiation environment. Measurement of mechanical properties of degraded pressure tubes is important for assessing their fitness for further operation. Presently as per safety guidelines imposed by the regulatory body, a few pre-decided pressure tubes are removed from the reactor core at regular intervals during the planned reactor shut down to carry out post irradiation examination (PIE) in a laboratory which consumes lots of man-rem and imposes economic penalties. Hence a system is indeed felt necessary which can carry out experimental trials for measurement of mechanical properties of pressure tubes under in situ conditions. The only way to accomplish this important objective is to develop a system based on an in situ measurement technique. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing an indentation test either on the outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ conditions. Considering the importance of such measurements, an In situ Property

  10. Direction of Heavy Water Projects

    1984-07-01

    Summary of the activities performed by the Heavy Water Projects Direction of the Argentine Atomic Energy Commission from 1950 to 1983. It covers: historical data; industrial plant (based on ammonia-hydrogen isotopic exchange); experimental plant (utilizing hydrogen sulfides-water process); Module-80 plant (2-3 tons per year experimental plant with national technology) and other related tasks on research and development (E.A.C.) [es

  11. Heavy water pumps; Pumpe D{sub 2}O

    Zecevic, V; Nikolic, M

    1963-12-15

    Continuous increase of radiation intensity was observed on all the elements in the heavy water system during first three years of RA reactor operation. The analysis of heavy water has shown the existence of radioactive cobalt. It was found that cobalt comes from stellite, cobalt based alloy which was used for coating of the heavy water pump discs in order to increase resistance to wearing. Cobalt was removed from the surfaces due to friction, and transferred by heavy water into the reactor where it has been irradiated for 29 876 MWh up to 8-15 Ci/g. Radioactive cobalt contaminated all the surfaces of aluminium and stainless steel parts. This report includes detailed description of heavy water pumps repair, exchange of stellite coated parts, decontamination of the heavy water system, distillation of heavy water. [Serbo-Croat] U toku prve tri godine eksploatacije reaktora RA uocen je neprekidni porast intenziteta zracenja na svim elementima u teskovodnom sistemu. Analizom teske vode utvrdjeno je postojanje radioaktivnog kobalta. Ustanovljeno je da kobalt potice od stelita, legure na bazi kobalta kojim su presvuceni rukavci vratila teskovodnih pumpi radi otpornosi na habanje. Kobalt je trenjem skidan sa povrsina, u toku rada prenosen je teskom vodom u reaktor i ozracivan u toku 29 876 MWh do specificne aktivnosti 8-15 Ci/g. Radioaktivni kobalt je kontaminirao sve povrsine od aluminijuma i nerdjajuceg celika. Ovaj izvestaj sadrzi detaljan opis remonta pumpi, zamene delova teskovodnih pumpi novim delovima bez stelitnog sloja, dekontaminacije teskovodnog sistema, destilacije teske vode.

  12. Study on the effect of moderator density reactivity for Kartini reactor

    Budi Rohman; Widarto

    2009-01-01

    One of important characteristics of water-cooled reactors is the change of reactivity due to change in the density of coolant or moderator. This parameter generally has negative value and it has significant role in preventing the excursion of power during operation. Many thermal-hydraulic codes for nuclear reactors require this parameter as the input to account for reactivity feedback due to increase in moderator voids and the subsequent decrease in moderator density during operation. Kartini reactor is cooled and moderated by water, therefore, it is essential to study the effect of the change in moderator density as well as to determine the value of void or moderator density reactivity coefficient in order to characterize its behavior resulting from the presence of vapor or change of moderator density during operation. Analysis by MCNP code shows that the reactivity of core is decreasing with the decrease in moderator density. The analysis estimates the void or moderator density reactivity coefficient for Kartini Reactor to be -2.17×10-4 Δρ/ % void . (author)

  13. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  14. Graphite moderated reactor for thermoelectric generation

    Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

    1998-01-01

    Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)

  15. Enhancement of efficacy of process water monitors in detecting heavy water leak in steam generator blow down lines

    Mitra, S.R.; Kohale, S.D.; Parida, B.K.; Gathe, G.D.; Pati, C.K.; Mudgal, B.K.; Niraj; Pawar, S.K.

    2006-01-01

    The Steam Generator (SG) serves as an interface between primary and secondary cycle in Pressurized Heavy Water Reactor (PHWR). Failure of steam generator tubes result in leaking of active heavy water in the secondary closed loop. In Tarapur Atomic Power Station-3 and 4 (TAPS- 3 and 4), Scintillator detectors are provided to detect on line heavy water leakages in SG and moderator heat exchangers by monitoring Nitrogen-16 ( 16 N) and Oxygen-19 ( 19 O) activities. Efficacy of detection of these activities at designed detector position on SG blow down line in presence of background radiation field is analysed theoretically. The count rate of 19 O and 16 N estimated at the detector position inside Reactor Building (RB) shows that detectors only respond to very high leak rates due to presence of high ambient radiation level even though sensitivity is appreciably good. For detector position in RB in the accessible areas and out side the RE containment, the travel time for the blow down feed water becomes moderately and very long respectively resulting in poor sensitivity. However the results show that wherever background levels is low, the efficacy of leak detection becomes considerably better than the results obtained when detector is placed inside RB. The study was validated during the reactor operation by recording the detector count rates due to prevalent ambient radiation level near to the detectors. Subsequently the detectors were relocated in an area inside RB where relocation was feasible, travel time of the blow down feed water was moderate and the area had an relatively low ambient radiation level. This paper discusses the methodology adopted during the study and results obtained during theoretical estimation and practical validation. (author)

  16. Solid methane cold moderator for the IBR-2 reactor

    Beliakov, A. A.; Tretiakov, I. T.; Shabalin, E. P.; Golikov, V. V.; Luschivkov, V I.

    1997-09-01

    The paper describes the research and design work carried out since 1986 at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research in Dubna to create a cryogenic moderator for the IBR-22 reactor using solid methane as a moderating substance.

  17. Zero energy reactor 'RB'

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  18. Reactor water level control device

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  19. The water chemistry of CANDU PHW reactors

    LeSurf, J.E.

    1978-01-01

    This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

  20. Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code

    Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu

    2014-01-01

    In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

  1. Heavy water. A production alternative for Venezuela

    A survey of heavy water production methods is made. Main facts about isotopic and distillation methods, reforming and coupling to a Hydrogen distillation plant are presented. A feasibility study on heavy water production in Venezuela is suggested

  2. Construction of the core of the 'heavy water-gas' reactor EL 4; Structures du coeur du reacteur 'eau- lourde-gaz EL 4'

    Bernard, J L; Foulquier, H; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The core of this reactor consists of a vessel containing heavy water, through which pass a series of pressure tubes for circulation of the cooling gas under boat pressure. The basic specifications which greatly influenced the design of this construction relate to aspects of safety in operation (fuel loading from both faces of the reactor, replacement of the components on both faces), neutronic demands (minimum absorption of the components lattice parameter, diameter of the pressure tubes) and thermal considerations (output temperature 500 C). These specifications have led to a' horizontal arrangement of the pressure tubes and raised very difficult problems of clearance, which make it impossible (for the dimensions of EL 4) to resort to expansion bellows on the pressure tubes. The result is a semi-rigid vessel in which the pressure tubes contribute to a large extent the mechanical resistance of the system by acting as a brace, whence the high stresses on the joints and pressure tubes (and the choice of zirconium alloys). The construction components include the pressure tube, the joints, the thermal insulation and the liner tube. A brief account is given of the testing methods used and the performances of these various units is particular. The safety factors foreseen for the pressure tube, and the design and manufacture, taking account of tolerances of the thickened ends necessary for fitting the tubes in place and designing the joints. The joints connecting the pressure tubes to the reactor tank, which are only accessible through the inside of the channel prolonging the pressure tube. These joints must not be a weak part in the construction. Two types have been developed: a rolled joint where the ends of the pressure tube are directly flanged onto the tank, and a welded joint using zircaloy-stainless steel transition pieces added to the ends of the pressure tube. All these joints are made by remote control and are removable. Two solutions have been found to the

  3. Water-immersion type ship reactor

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

    In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

  4. Boiling water reactor

    Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

    1975-01-01

    Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

  5. The purification by ion exchange resins of the heavy water la the reactors EL1 and EL2. B - study of the general properties of the resins used

    Fourre; Platzer

    1957-01-01

    Within the programme of the pile heavy water purification project, organized by the stable Isotopes Section, we have carried out a certain number of tests on ion exchange resins. The problem posed by the stable Isotopes Section was to determine the conditions of utilisation of ion exchange resins, knowing that they would be employed in a system branching off the heavy water circuit in the piles. These investigations were carried out in close collaboration with the stable Isotopes Section, and were guided chiefly by the extremely short delay permitted between the laboratory study and its application to the piles. The tests are divided into two groups: 1- General properties of the resins. 2- Utilisation of the resins, particularly in an apparatus similar to those mounted on the piles but of smaller dimensions. (author) [fr

  6. Study of new structures adapted to gas-graphite and gas-heavy water reactors; Etude de structures nouvelles adaptees aux reacteurs graphite-gaz et eau lourde-gaz

    Martin, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [French] L'experience acquise par l'exploitation des reacteurs de MARCOULE, la construction et le demarrage des reacteurs d

  7. Fuel and heavy water availability

    1980-01-01

    The general guidelines for the Working Group's evaluation of the availability of nuclear fuel and heavy water were set at the Organizing Conference of the International Nuclear Fuel Cycle Evaluation (INFCE), which was held in Washington, United States of America, 19-21 October 1977. The agreed technical and economic scope for the evaluation was to: (1) Estimate needs for nuclear energy and correlated needs for uranium and heavy water according to different fuel cycle strategies; (2) Review uranium availability with specific regard to: Assessment of resources and production capacities; policies and incentives for encouraging exploration and production including joint ventures; marketing policies and/or guarantees of sales for companies investing in exploration and production; marketing policies and/or guarantees of supply for utilities; technical development of exploration, mining and milling methods; (3) Review heavy water availability; (4) Review thorium availability; (5) Consider special needs of developing countries. The illustrations of availability and requirements developed in this report do provide a useful framework for considering future options and alternatives for the development of nuclear power

  8. Various analytical techniques used for the measurement of isotopic purity of heavy water at Madras Atomic Power Station

    Satyanarayanan, V.; Umapathy, P.; Bhaskaran, R.; Nagarajan, J.; Pradeep, Jeena; Ayyar, S.R.

    2008-01-01

    The paper deals with the various techniques used for the measurement of isotopic purity of heavy water samples received from different sources viz. reactor systems, heavy water upgrading plant and fresh consignment from heavy water production plants. Heavy water is used in PHWRs as moderator and primary coolant. Isotopic Purity is an important parameter to be monitored/analysed regularly for both the systems. There is a minimum isotopic purity level to be maintained in the moderator system due to neutron economy/fuel burnup and in the case of coolant system the measurement is of paramount importance due to its safety considerations. The selection of the method of analysis depends on the isotopic range. The techniques used to measure the isotopic purity of heavy water are a) Infrared Spectrophotometry b) Refractometry c) Densitometry. Infrared spectrometer uses the property of molecular absorption of IR radiation by HOD species and the absorbance is the measure of isotopic purity. This technique is generally used for measuring high isotopic (80-99.98%) and low isotopic samples. Refractometer uses the property of refractive index of heavy water. The difference in refractive indices of light water and heavy water is 0.0048. A 1 % change in D 2 O concentration would thus equal to 0.000048 refractive index units. This method is used for determining the approximate isotopic value of a sample. Density meter uses the property of difference in densities of light and heavy water. The difference in density of 99.999% D 2 O and light water is 0.107540 which covers the whole range of interest. The experience gained with these techniques in the measurements of isotopic purity of various samples are presented in this paper. (author)