Investigation of Quench-16 Experiment With MELCOR Computer Code

This paper presents a comparison of MELCOR calculated results with experimental data for the QUENCH-16 experiment. The analysis for the air ingress experiment QUENCH-16 has been performed by INRNE. The calculations have been performed with MELCOR code. The QUENCH-16 experiment has been performed on 27-th of July 2011 in the frame of the EC-supported LACOMECO program. The experiments have focused on air ingress investigation into an overheated core following earlier partial oxidation in steam. QUENCH-16 has been performed with limited pre-oxidation and low air flow rate. One of the main objectives of QUENCH-16 was to examine the interaction between nitrogen and oxidized cladding during a prolonged period of oxygen starvation. The bundle is made from 20 heated fuel rod simulators arranged in two concentric rings and one unheated central fuel rod simulator, each about 2.5 m long. The tungsten heaters were surrounded by annular ZrO2 pellets to simulate the UO2 fuel. The geometry and most other bundle components are prototypical for Western-type PWRs. To improve the obtained results it has been made a series of calculations to select an appropriate initial temperature of the oxidation of the fuel bundle and modified correlation oxidation of Zircaloy with MELCOR computer code. The compared results have shown good agreement of calculated hydrogen and oxygen starvation in comparison with test data.

Power Cycles of Generation III and III+ Nuclear Power Plants

It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living. In general, electrical energy can be generated by: 1) non-renewable-energy sources such as coal, natural gas, oil, and nuclear; and 2) renewable-energy sources such as hydro, wind, solar, biomass, geothermal and marine. However, the main sources for electrical-energy generation are: 1) thermal - primary coal and secondary natural gas; 2) “large” hydro and 3) nuclear. The rest of the energy sources might have visible impact just in some countries. Modern advanced thermal power plants have reached very high thermal efficiencies (55–62%). In spite of that they are still the largest emitters of carbon dioxide into atmosphere. Due to that, reliable non-fossil-fuel energy generation, such as nuclear power, becomes more and more attractive. However, current Nuclear Power Plants (NPPs) are way behind by thermal efficiency (30–42%) compared to that of advanced thermal power plants. Therefore, it is important to consider various ways to enhance thermal efficiency of NPPs. The paper presents comparison of thermodynamic cycles and layouts of modern NPPs and discusses ways to improve their thermal efficiencies.

Thermal Hydraulic Studies of a Fluoride Salt Cooled High Temperature Test Reactor With Different CFD Methods

The Fluoride-salt-cooled High temperature Reactor (FHR) is new reactor concept-about a decade old which is mainly on going in China and U.S. The preliminary thermal-hydraulic studies of the Fluoride salt cooled High temperature Test Reactor (FHTR) is necessary for the development of the FHR technology. In this paper, the thermal-hydraulics of FHTR (also called TMSR-SF) designed by Shanghai Instituted of Applied Physics (SINAP) is studied in different power modes. The temperature distributions of the coolant and the fuel pebble are obtained using a steady-state thermal-hydraulic analysis code for FHR. The comprehensive local flow and heat transfer are investigated by computational fluid dynamics (CFD) for the locations where may have the maximum pebble temperature based on the results from single channel analysis. The profiles of temperature, velocity, pressure and Nu of the coolant on the surface of the pebble as well as the temperature distribution of a fuel pebble are obtained and analyzed. Numerical results showed that the results of 3-D simulation are in reasonable agreement with that of single channel model and also illustrated safety operation of the preliminary designed TMSR-SF in different power mode.

Cross-Comparison of One-Dimensional Thermal-Hydraulic Codes on Natural Circulation Analysis of NACIE Loop Test for Lead-Alloy Cooled Advanced Nuclear Energy Systems (LACANES)

As one of the Generation-IV reactor concepts, lead-alloy-cooled advanced nuclear energy systems (LACANES) have been studied worldwide in order to utilize the advantages of good heat transfer properties, neutron transparency and chemical inertness with air and water. Since the Fukushima accident, the passive safety aspect of the LACANES is increasingly emphasized due to outstanding natural circulation capability. To investigate the thermal-hydraulic capability of LBE, an international cooperation has been performed under OECD/NEA program, under the guidance of the Nuclear Science Committee by a task force named as Lead Alloy Cooled Advanced Nuclear Energy Systems (LACANES) since 2007. This international collaboration had dealt with computational benchmarking of isothermal LBE forced convection tests in the phase I, and the working group published a guideline for using one-dimensional system codes to simulate LBE forced circulation test results from HELIOS loop. The phase II was started after that, to give an additional guideline in the case of natural circulation. NACIE, one of benchmarking targets for the phase II which is a rectangular-shape loop located at ENEA-Brasimone Research Centre, Italy. NACIE test results were benchmarked by each participant using their one-dimensional thermal-hydraulic codes, and they are to follow the guideline from the LACANES phase I for regions where hydraulic loss occurs. Due to the selection of hydraulic loss coefficient relations by users, the cross-comparison results of international participants showed some discrepancies and the estimated mass flow rates had 13% of maximum error. Also, the future R&D areas are identified.

Burnup Credit in the Criticality Safety Analysis of Spent Fuel in Storage Systems

In this paper accounting of spent nuclear fuel (SNF) burnup of RBMK-1000 with actinides and full isotopic composition has been performed. The following characteristics were analyzed: initial fuel enrichment, burnup fraction, axial burnup profile in the fuel assembly (FA) and fuel weight. As the results show, in the first 400 hours after stopping the reactor, there is an increase in the effective neutron multiplication factor (keff) due to beta decay of 239Np into 239Pu. Further, from 5 to 50 years, there is a decrease in keff due to beta decay of 241Pu into 241Am. Beyond 50 years there is a slight change in the criticality of the system. Accounting for nuclear fuel burnup in the justification of nuclear safety of SNF systems will provide an opportunity to increase the volume of loaded fuel and thus significantly reduce technology costs of handling of SNF.

Accident at Fukushima Dai-Ichi Nuclear Power Plant: Lessons Learned for the Czech Republic

Accident at Fukushima Dai-Ichi nuclear power plant significantly affected the nuclear industry at time when everybody was expecting the so called nuclear renaissance. There is no question that the accident has at least slowed it down. Research into this accident is taking place all over the world. In this paper we present the findings of research on Fukushima nuclear power plant accident in relation to the Czech Republic. The paper focuses on the analysis of human performance during the accident. Lessons learned from the accident and main human errors are presented. First the brief factors affecting the human performance are discussed. They are followed by the short description of activities on units 1–3. The key human errors in the accident mitigation are then identified. On unit 1 the main error is wrong understanding and operation of isolation condenser. On unit 2 the main errors were unsuccessful depressurization with subsequent delay of coolant injection. On unit 3 the main error is the shutdown of high pressure cooling injection system without first confirming that different means of cooling are available. These errors lead to fuel damage. On unit 1 the fuel damage was probably impossible to prevent, however on unit 2 and 3 it could be probably prevented. The lessons learned for the Czech Republic were presented. They can be summarizes as follows: be sure that plant personnel can and knows how to monitor and operate the crucial plant components, be sure that the procedures on how to fulfill the critical safety functions are available in the symptomatic manner for situations when there is no power available at the plant, train personnel for these situations and have sufficient human resource available for these situations.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

High Efficiency Gamma-Beta Blind Alpha Spectrometry for Nuclear Energy Applications

A novel, Centrifugally Tensioned Metastable Fluid Detector (CTMFD) sensor technology has been developed over the last decade to demonstrate high selective sensitivity and detection efficiency to various forms of radiation for wide-ranging conditions (e.g., power level, safeguards, security, and health physics) relevant to the nuclear energy industry. The CTMFD operates by tensioning a liquid with centrifugal force to weaken the bonds in the liquid to the point whereby even a femto-scale nuclear particle interactions can break the fluid and cause a detectable vaporization cascade. The operating principle has only peripheral similarity to the superheated bubble chamber based superheated droplet detectors (SDDs); instead, CTMFDs utilize mechanical “tension pressure” instead of thermal superheat offering a lot of practical advantages. CTMFDs have been used to detect a variety of alpha and neutron emitting sources in near real-time. The CTMFD is selectively blind to gamma photons and betas allowing for detection of alphas and neutrons in extreme gamma/beta background environments such as spent fuel reprocessing plants or under full power conditions within an operating nuclear reactor itself. The selective sensitivity allows for differentiation between alpha emitters including the isotopes of Plutonium. Mixtures of Plutonium isotopes have been measured in ratios of 1:1, 2:1, and 3:1 Pu-238:Pu-239 with successful differentiation. Due to the lack of gamma-beta background interference, the CTMFD’s LLD can be effectively reduced to zero and hence, is inherently more sensitive than scintillation based alpha spectrometers or SDDs and has been proven capable to detect below femtogram quantities of Plutonium-238. Plutonium is also easily distinguishable from Neptunium making it easy to measure the Plutonium concentration in the NPEX stream of a UREX reprocessing facility. The CTMFD has been calibrated for alphas from Americium (5.5 MeV) and Curium (∼6 MeV) as well. The CTMFD has furthermore, recently also been used to detect spontaneous and induced fission events which can be differentiated from alpha decay allowing for detection of fissionable material in a mixture of isotopes. This paper discusses these transformational developments which are also being entered for real-world commercial use.

Validation of a New PURC Module for Calculating Probability Table in Unresolved Resonance Region in RXSP Code

Taking into consideration the unresolved resonance self-shielding effect, probability table is one of the most important and natural methods used to simulate the neutron transport in unresolved resonance range in reactor physics. A module named PURC for probabilistic unresolved resonance calculation which is embedded in the RXSP Beta2.0 code, has been developed by REAL team in Department of Engineering Physics in Tsinghua University. It is the optimization in sort algorithm that make PURC module more efficient. After applying the OpenMP parallel algorithm into this module, it has improved computational efficiency by more than one order of magnitude comparing with the corresponding functional module named PURR in NJOY code. Meanwhile the computational accuracy of PURC module is validated and verified by series of microscopic cross section comparisons and macroscopic criticality benchmarks.

Structural Supportive Software Tool for Fire PSA

Continually increasing requirements on nowadays full scope PSA L1 and L2 as whole, which is multiplied by importance of specific data for all modes of operation of nuclear power plant, highlight role of input data used in PSA quantification process. This fact also emphasizes the role of capability to process all necessary information to analyze all nuclear plant modes by appropriate way. Even if abovementioned aspects are relevant for all parts of nowadays PSAs, their importance is critical for internal hazards including specific fire analysis. Because internal fire analysis forms one of the most challenging PSA tasks, requiring interdisciplinary work including processing and integration of extensive amount of data in such a way that fire analysis results are fully consistent with internal PSA events and can be directly incorporated into PSA project. Application of tailored information system forms one of the ways to speed up analyzing process, enhances manageability and maintainability of particular PSA projects and provides effective reporting mean to document process of work as well as traceable and human readable documentation for customers. Such information system also allows implementing rapid changes in processing input data and reduces the risk of human error. Usage of information systems for modification of input data for Living PSA is invaluable. Transparent highly automatized processing of input data allows the analyst to obtain more accurate and better insight to evaluate aspects of particular fire and its consequences. This paper provides brief overview of VUJE approach and experience in this area. The paper introduces general purpose of database developed for PSA needs containing data for relevant PSA structure system and components as well as information relevant for flood and fire analyses. Paper explains as this basic data source is enhanced by adding several relatively independent tiers to employ all common data for fire PSA purpose. Paper also briefly introduces capability of such system to generate integrated documentation covering all stages of fire analyses, covering all screening stages of fire analysis as well as future plans to enhance this part of work in such a way to be capable to build automatic interface between PSA model and fire database to enable PSA model parameters automatic updating and expansion of fires in combinations of initiating events (for example Fire and seismic event).