Small Modular Reactor and Large Nuclear Reactor Fuel Cost Comparison

2014 ◽  
Author(s):  
Christopher P. Pannier ◽  
Radek Škoda
Keyword(s):  
Nuclear Reactor ◽  
Cost Comparison ◽  
Design Parameters ◽  
Fuel Cost ◽  
Uranium Enrichment ◽  
Pressurized Water ◽  
Nuclear Reactor Fuel ◽  
Water Reactors ◽  
Near Term ◽  
Fuel Costs

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.

scholarly journals An Overview of the Pressurized Thermal Shock Issue in the Context of the NURESIM Project

2009 ◽  
Vol 2009 ◽  
pp. 1-13 ◽  
Author(s):  
D. Lucas ◽  
D. Bestion ◽  
E. Bodèle ◽  
P. Coste ◽  
M. Scheuerer ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Cfd Simulation ◽  
Two Phase ◽  
Pressurized Water ◽  
Simulation Tools ◽  
Pressurized Thermal Shock ◽  
Qualitative Level ◽  
Water Reactors ◽  
Near Term ◽  
Accident Scenarios

Within the European Integrated Project NURESIM, the simulation of PTS is investigated. Some accident scenarios for Pressurized Water Reactors may cause Emergency Core Coolant injection into the cold leg leading to PTS situations. They imply the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. This paper focuses on two-phase conditions that are potentially at the origin of PTS. It summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor,that is, the cold leg and the downcomer, where the “PTS fluid-dynamics" is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two-phase flow subphenomena are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance is still limited. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet well controlled, leading to slightly conservative predictions.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

How to Simulate the Microstructure Induced by a Nuclear Reactor with an Ion Beam Facility : DART

2009 ◽  
Vol 1215 ◽  
Author(s):  
Laurence Luneville ◽  
David Simeone ◽  
Gianguido Baldinozzi ◽  
Dominique Gosset ◽  
yves serruys
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Ion Beam ◽  
Fast Neutrons ◽  
Relaxation Processes ◽  
Pressurized Water ◽  
Water Reactors ◽  
Displacement Per Atom ◽  
The Impact

AbstractEven if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and ion beam facilities on a defined material. In this paper, a formalism is presented to evaluate the displacement cross-sections pointing out the effect of the anisotropy of nuclear reactions. From this formalism, the impact of fast neutrons (with a kinetic energy En superior to 1 MeV) is accurately described. This point allows calculating accurately the displacement per atom rates as well as primary and weighted recoil spectra. Such spectra provide useful information to select masses and energies of ions to perform realistic experiments in ion beam facilities.

scholarly journals Nuclear viewpoint in India

2017 ◽  
Vol 4 ◽  
Author(s):  
Anshu Bharadwaj ◽  
Lakshminarayana Venkat Krishnan ◽  
Subramaniam Rajagopal
Keyword(s):  
Heavy Water ◽  
Nuclear Power ◽  
Clean Energy ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Water Reactors ◽  
Near Term ◽  
Enriched Uranium ◽  
Installed Capacity

ABSTRACTNuclear power is a crucial source of clean energy for India. In the near-term, India is focusing on thermal reactors using natural and enriched uranium. In the long-term, India is exploring various options to use its large thorium reserves.India’s present nuclear installed capacity is 5680 MW, which contributes to about 3.4% of the annual electricity generation. However, nuclear power is an important source of energy in India’s aspirations for energy security and also in achieving its Intended Nationally Determined Contributions (INDC), of 40% fossil free electricity, by 2030. India has limited uranium reserves, but abundant thorium reserves. The Nuclear Suppliers Group (NSG) lifted restrictions on trade with India, in 2008, enabling India to import uranium (natural and enriched) and nuclear reactors. In the near–term (2030), the nuclear capacity could increase to about 42,000 MW. This would be from a combination of domestic Pressurized Heavy Water Reactors (PHWR) and imported Pressurized Water Reactors (PWR). For the long–term (2050), India is exploring various options for utilising its vast thorium reserves. This includes Advanced Heavy Water Reactor and Molten Salt Breeder Reactor. However, generating public acceptance will be crucial to the expansion of the nuclear power program.

scholarly journals Anthropogenic Radiocarbon: Past, Present, and Future

Radiocarbon ◽  
1986 ◽  
Vol 28 (2A) ◽  
pp. 668-672 ◽  
Author(s):  
Pavel Povinec ◽  
Martin Chudý ◽  
Alexander Šivo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Pressurized Water Reactors ◽  
Anthropogenic Radionuclides ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Weapon Testing ◽  
Water Reactors

14C is one of the most important anthropogenic radionuclides released to the environment by human activities. Weapon testing raised the 14C concentration in the atmosphere and biosphere to +100% above the natural level. This excess of atmospheric C at present decreases with a half-life of ca 7 years. Recently, a new source of artificially produced 14C in nuclear reactors has become important. Since 1967, the Bratislava 14C laboratory has been measuring 14C in atmospheric 14CO2 and in a variety of biospheric samples in densely populated areas and in areas close to nuclear power plants. We have been able to identify a heavy-water reactor and the pressurized water reactors as sources of anthropogenic 14C. 14C concentrations show typical seasonal variations. These data are supported by measurements of 3H and 85Kr in the same locations. Results of calculations of future levels of anthropogenic 14C in the environment due to increasing nuclear reactor installations are presented.

An Evaluation of Post-LOCA Debris on Multi-Stage Safety-Related Pumps in PWR ECCS in Support of Generic Safety Issue 191 Closure

2008 ◽  
Author(s):  
L. Ike Ezekoye ◽  
Rolv Hundal ◽  
Paul V. Pyle
Keyword(s):  
Nuclear Reactor ◽  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
Design Parameters ◽  
Two Phase ◽  
Pressurized Water ◽  
Pump Design ◽  
Critical Dimensions ◽  
Multi Stage ◽  
Core Cooling

The U.S. Nuclear Regulatory Commission (NRC) issued Generic Safety Issue (GSI) 191 covering the ability of nuclear reactor containment building (RCB) sumps to support long-term core cooling post-accident for Pressurized Water Reactors (PWRs). The issue is that a postulated Loss-of-Coolant Accident (LOCA) for a PWR would result in the initial escape of high-pressure, high-temperature subcooled coolant from the pipe break location in the form of a two-phase jet. The impingement of the jet may damage materials used inside the RCB, resulting in the generation of debris. The debris transported to and through the sump screen when the Emergency Core Cooling (ECC) and the Containment Spray (CS) Systems are realigned to draw suction from the containment sump will be ingested by pumps in the flow path. Since wear affects the dynamics of the pump and hence vibration, the extent to which any pump can survive the abrasive nature of the ingested debris depends on how well one can predict the wear of pump critical dimensions using the debris mix and pump design parameters. This paper describes the issues to be considered in pump assessment and presents wear models that can be used to assess pump operability and performance. It shows that depending on assumptions made relative to the wear pattern, different results can be reached using the wear models. The question of what makes sense is discussed.

scholarly journals Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

2012 ◽  
Vol 66 (3) ◽  
pp. 291-299 ◽  
Author(s):  
Grégory Lefèvre ◽  
Ljiljana Zivkovic ◽  
Anne Jaubertie
Keyword(s):  
Nuclear Reactor ◽  
Primary Circuit ◽  
Particle Removal ◽  
Mechanical Seals ◽  
Solution Ph ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Key Factor ◽  
Water Reactors ◽  
Chemical Conditions

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition) in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek) theory and used as such to interpret this industrial phenomenon.

A Systematic Review of PCTRAN-Based Pressurized Water Reactor Transient Analysis

2021 ◽  
Author(s):  
Suubi Racheal ◽  
Yongkuo Liu ◽  
Miyombo Ernest ◽  
Abiodun Ayodeji
Keyword(s):  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Research Education ◽  
User Expertise ◽  
Scale Down ◽  
Water Reactors ◽  
The Impact ◽  
And Training

Abstract The impact of nuclear accidents has been a topic of debate since the construction of the first nuclear reactor, and still stands as a key issue of public concern. Several codes and simulators have been used to study the transient progression in pressurized water reactors, and to evaluate the technical measures adopted to scale down the risk of accidents. However, some of these codes are not suitable for multipurpose research and training as they require significant user expertise, leading to analysis uncertainties largely from the code user effect. This paper presents a bird-eye view of one of the most widely used nuclear reactor transient analyzer — the Personal Computer Transient Analyzer (PCTRAN). This paper discusses the comparative advantages of the simulator from the users’ perspective, with specific attention to its utilization both for research and training. The paper also demonstrates the ease of usage by simulating common transient in a pressurized water reactor. Finally, observations and possible improvements to the code to increase its usability in research, education and training are discussed. This work aims to evaluate the robustness of the simulator towards better utilization for research and training, especially in nuclear newcomer countries.

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