Reactors for low-temperature nuclear heat supply

Nuclear sources are not only covering more than 16% of today's electricity production but can also supply heat for district heating and industrial needs. Thus the nuclear generated heat substitutes for fossil fuels with good efficiency and economy and with much higher environmental cleanliness. Low-temperature nuclear heat is gained in several countries from the reactors of nuclear power plants by co-generation of heat and electricity which is already a proven technology. Specialized nuclear heating plants are in an early stage of development. The paper gives an overview of the situation worldwide and shows also specific common safety characteristics of these reactors.     

Design concept of self-contained low power reactor master for heat supply

This paper demonstrates technical features and conceptual scheme of innovative self-contained low power reactor MASTER for heat supply. Neutron-physical and thermo-hydraulic characteristics of this reactor are analyzed. The possibility of power self-control and minimization of reactivity swing during fuel burnup are considered.     

Economic feasibility of heat supply from simple and safe nuclear plants

Use of nuclear energy as a heating source is greatly challenged by the economic factor since the nuclear heating reactors have relative small size and often the lower plant load factor. However, use of very simple reactor could be a possible way to economically supply heat. A deep pool reactor (DPR) has been designed for this purpose. The DPR is a novel design of pool type reactor for heat only supply. The reactor core is put in a deep pool. By only putting light static water pressure on the core coolant, the DPR will be able to meet the temperature requirements of heat supply for district heating. The feature of simplicity and safety of DPR makes a decrease of investment cost compared to other reactors for heating only purposes. According to the economical assessments, the capital investment to build a DPR plant is much less than that of a pressurized reactor with pressure vessels. For the DPR with 120 or 200 MW output, it can bear the economical comparison with a usual coal-fired heating plant. Some special means taken in DPR design make an increase of the burn-up level of spent fuel and a decrease of fuel cost. The feasibility studies of DPR in some cities in China show that heating cost using nuclear energy is only one third of that by coal and only one tenth of that by nature gas. Therefore, the DPR nuclear heating system provides an economically attractive solution to satisfy the demands of district heating without contributing to increasing greenhouse gas emissions.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

Inherently safe pool-type reactor as a generator of low-grade heat for district heating, air conditioning and salt water desalination

The society has a heavy demand for low-grade heat to satisfy its various needs. Different factors govern the expediency of applying nuclear reactors for these purposes. The required capacity of heat sources varies in a very wide range. In a majority of cases heat sources have to be located in the immediate vicinity of the users, therefore, nuclear reactors to be used for heat generation must feature enhanced safety. Pool-type reactors can be successfully used for producing low-grade heat. Owing to their design they feature a very high safety level. The absence of positive pressure excludes the possibility of a sudden rupture of reactor tank (vessel) or a fast loss of coolant. The availability of a large amount of water in the tank ensures long-term accumulation of residual heat. The adopted integral layout of equipment, as well as natural circulation of primary coolant improve reactor reliability and safety even further. Negative temperature coefficients of reactivity provide for reactor self-protection against reactivity accidents. Pool-type reactors can be used in newly established heat supply systems and can be built in the operation systems as well, which allows to reduce fossil fuel consumption by 80–90% depending on local conditions. Pool-type reactor heat can be used for desalinating salt water and for cooling water in absorption refrigerating machines with subsequent utilization of cold water for air conditioning, cooling of special premises, and the like. Pool-type reactors can also generate electric power to their in-house needs as well as household power requirements of a neighboring town.     

A 10 MWe pressurized water reactor for safe and reliable power

A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve “walk-away” safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control.     

Nonlinear statics for a large reactor with deep fuel depletion

Translated from Atomnaya Énergiya, Vol. 64, No. 3, pp. 165–170, March, 1988.     

钠冷快堆中熔融池模型的建立与验证

快堆在超设计基准事故下运行时,会导致钠沸腾和干涸,如果不能及时停堆,接着就会产生燃料元件的熔化坍塌,在组件盒下部形成熔融池.为了对熔融池给出合理的安全分析,采用机理建模的方法,建立了完整的熔融池模型,并在法国的SCARABEE系列实验中的BF1三种功率的实验上进行了验证,和实验吻合较好,通过和所验证过的GEYSER及BF2等实验模型进行比较,得出了有关熔融池机理的相关结论.通过排热和温升等相关数据的比较,对熔融池向外的排热形式给出了合理分析,并得出了相关结论.     

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility.Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time.For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the U.S. Nuclear Regulatory Commission (NRC) and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.     

An integrated safe shutdown heat removal system for light water reactors

A design concept is introduced for an integrated safe shutdown heat removal system (ISSS) for light water reactors that is independent of all components and systems outside the primary containment (other than the elements of the ISSS itself) and which includes an integral stored water supply. The principal purpose is to obtain a simple, reliable, and highly protected means for water makeup and steam release for basic post-scram cooling. The objective of the system is to provide safe shutdown cooling under virtually all emergency conditions which do not involve the loss of coolant caused by piping failures within the primary containment. Thus, the system is designed for the hazards of fire and sabotage as well as numerous other potential accidents, all of them considered as a set of events to be handled by the same system, rather than as discrete problems — each with its own, sometimes contradictory, solution.     

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper (Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.     

The design and experimental validation of an emergency core cooling system for a pool type research reactor

This paper presents the design of the Emergency Core Cooling System (ECCS) for the IEA-R1m pool type research reactor. This system with passive features, uses sprays installed above the core. The experimental program performed to define system parameters and to demonstrate to the licensing authorities, that the fuel elements limiting temperature is not exceeded, is also presented. Flow distribution experiments using a core mock-up in full-scale were performed to define the spray header geometry and spray nozzles specifications as well as the system total flow rate. Another set of experiments using electrically heated plates simulating heat fluxes corresponding to the decay heat curve after full power operation at 5 MW was conducted to measure the temperature distribution at the most critical position. The observed water flow pattern through the plates has a very peculiar behavior resulting in a temperature distribution which was modelled by a 2D energy equation numerical solution. In all tested conditions the measured temperatures were shown to be below the limiting value.     

The encapsulated nuclear heat source reactor for low-waste proliferation-resistant nuclear energy

The encapsulated nuclear heat source (ENHS) is a new Pb-Bi cooled modular reactor concept that features a combination of the following useful features that may make nuclear energy more attractive: (1) 20 years of full power operation without refueling. (2) Nearly constant fissile fuel contents and k_eff. (3) No on-site refueling and fueling hardware. (4) The ENHS modules are factory manufactured and transported already fueled to the site. (5) No access to neutrons. (6) No mechanical connections between the ENHS module and the energy conversion plant (The ENHS module has the function of a nuclear battery — with 20 years of full power operation at 125 MW_th). (7) At end of life, the ENHS module serves as a spent fuel storage cask and, later, as a spent fuel shipping cask. That is, the fuel is locked inside the ENHS from “cradle to grave”. (8) 100% natural circulation resulting in passive load following capability and autonomous control. This combination of features offers a highly safe nuclear energy system that is characterized by low waste, high proliferation resistance and high uranium utilization. The low waste and high uranium ore utilization are achieved by recycling the Pu and MA many times using a proliferation-resistant dry process; only fission products are to be extracted between cycles. Spent LWR fuel can provide for the HM make-up. The high level of proliferation resistance is obtained by restricting access to the fuel and neutrons and by eliminating the economic incentive of the client country to invest in sensitive technologies or infrastructure that can be used for clandestine production of strategic nuclear materials.     

Design evaluation of a small high-temperature reactor for process heat applications

The high-temperature reactor is also suitable for process heat application, in particular in smaller size units. On the basis of the results of R & D and demonstration work on coal refinement some improvements for small high-temperature reactor for process heat applications are discussed. These are: increase of the gas outlet temperature, the gas inlet temperature, the decrease of the overall system pressure, and non-integration of the primary circuit components. Some of these design evaluations are based on the results of the pre-project for the extension of the AVR-reactor in Jülich to be used as a process heat plant.     

Thermal-hydraulic modeling and analysis of a tank in pool reactor for normal operation and loss of flow transient

A transient thermal-hydraulic model entitled Tank in Pool Reactor Thermal-Hydraulic Analysis (TPRTHA) has been developed to simulate the steady-state operation and loss of flow transient for a tank in pool type research reactor. The model solves the momentum equation, energy equation and general conduction equation in cylindrical coordinates in order to predict the coolant velocity, coolant temperature and fuel rod temperature distribution respectively. The analytical solution is utilized for steady-state calculation while the finite difference technique with implicit scheme is adopted for transient calculation. The model divides the active core into a specified axial regions and the fuel rod into a specified radial zones, then a nodal calculation is performed for both average and hottest rods with a chopped cosine shaped heat generation flux. The model also predicts the heat flux leading to onset of nucleate boiling and the critical heat flux as well as the flow inversion phenomenon. The model is used to simulate a 2 MW reactor with downward flow direction and different types of fuel bundles of different powers and different flow rates. The best-estimate thermal-hydraulic safety margins are determined and the model results are analyzed and discussed.     

On heat transfer characteristics of real and simulant melt pool experiments

This paper presents results of analytical studies on natural convection heat transfer in scaled and/or simulant melt pool experiments related to the pressurized water reactor in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing the relevant physical processes in prototypical situations is examined. Experiments and experimental approaches are analyzed by focusing on their ability to represent prototypical situations. Calculations are performed to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. In the present analysis, Rayleigh numbers are limited to 10¹², where uncertainties in turbulence modelling do not override other uncertainties. Calculations are performed to explore limitations of using side wall heating and direct electrical heating. The need for further experimental and analytical efforts is also discussed.