Novel Non-Evaporable Getter Materials and Their Possible Use in Fusion Application for Tritium Recovery

Non-evaporable getters (NEGs) are metallic compounds of the IV group, particularly titanium and/or zirconium-based alloys and are usually used as pumps in vacuum technologies since they are able to sorb, by chemical reactions, most of the active gas molecules, with particular efficacy towards hydrogen isotopes. This work suggests an alternative application of these materials to fusion nuclear reactors, where there is the need to recover small amount of tritium from the large helium flow rate composing the primary coolant loop. Starting from the tritium mass balance inside the primary coolant loop, the amount of coolant to be routed inside the coolant purification system (CPS) is identified. Then a feasibility study, based on the bulk getter theory, is presented by considering three different commercial alloys, named ST707, ST101 and ZAO. The results provide the mass, the area and the regeneration parameters of the three different alloys necessary to fulfill the requirements of the CPS unit. By comparing the features of the three alloys, the ZAO material appears the most promising for the proposed application because it requires the lower amount of material and a lower number of regeneration cycles.

Buildup and Decay Analysis of Corrosion Products Activity in Primary Coolant Loop of AP-1000

The broad half-life range of Activated Corrosion Products (ACPs) results in major radiation exposure throughout reactor operation and shutdown. The movement of unpredicted activity hot spots in coolant loop can bring about huge financial and dosimetric impacts. The PWR operating experience depicts that activity released during reactor operation and shutdown cannot be estimated through a simple correlation. This paper seeks to analyze buildup and decay behavior of ACPs in primary coolant loop of AP-1000 under normal operation, power regulation and shutdown modes. The application of a well-tested mathematical model is extended in an in-house developed code CPA-AP1000, to simulate the behavior of dominant Corrosion Products (CPs), by programing in MATLAB. The MCNP code is used as a subroutine of the program to model the reactor core and execute energy dependent neutron flux calculations. It is observed that short-lived CPs (56Mn, 24Na) build up rapidly under normal operation mode and decay quickly after the reactor is shutdown. The long-lived CPs (59Fe, 60Co, 99Mo) have exhibited slow buildup under normal operating conditions and likewise sluggish decay after the shutdown. To analyze activity response during reactor control regime, operating power level is promptly decreased and in response specific activity of CPs also followed decreasing trend. It is noticed that activity of CPs drops slowly during reactor control regime in comparison to emergency scram. The results are helpful in estimating radiation exposure caused by ACPs during accessibility of the equipment in coolant loop, under normal operation, power regulation and shutdown modes. Moreover, current analyses provide baseline data for further investigations on ACPs in AP-1000, being a new reactor design.

Study on the Shaking Dynamic Response of Steam Generator LOCA

Dynamic response research of steam generator loss of coolant accident (SG LOCA) is essential for the reliability and safety consideration. According to the differences of LOCA loading phenomena, two types of LOCA loads affect the SG: rarefaction wave travels through the primary fluid in the U-tubes, and the SG shakes due to reactor coolant loop(RCL) motions transmitted by the primary loop piping, former loading phenomena evaluation is called SG rarefaction analysis while latter is called shaking analysis. This paper place particular emphasis on shaking analysis. At present, the published literatures about LOCA mainly focus on RCL LOCA, reactor LOCA and fuel assembly LOCA, few reports concentrate on shaking dynamic response analysis of SG LOCA. Both Westinghouse and AREVA’s methods according to their research reports are to decouple the SG from the RCL: This method results in low computational efficiency as RCL LOCA and SG LOCA are evaluated separately and the decoupling error is uncertainty, meanwhile, in the vicinity of the nodes where the displacements are imposed, distorted reaction forces are usually found. Through reasonable simplification and equivalence, a detailed nonlinear FEM model of steam generator (SG) of a China 3rd generation nuclear power plant (NPP) is established, this model is then connected with the reactor coolant loop (RCL) to carry out the SG LOCA shaking dynamic response analysis. By calculation, the maximum absolute stresses of SG heat transfer tube bundle and its variation with tube diameter and upper supports reacting forces are obtained. In order to study the effect of SG decoupling from the RCL on shaking dynamic response, a comparative study of decoupling /coupling methods is conducted. Results shows that SG decoupling has a significant impact on the calculation result, the calculation method of coupling is more closer to the real situation and worthy to recommended. Related analytical procedures and calculation results lay the foundation for future SG shaking dynamic analysis and SG design of subsequent power plants.

On-Site Vibration Test for Measuring Damping Ratio of PWR Reactor Coolant Loop

In the Application for Construction Plan License after the Great East Japan Earthquake, it was needed to revalidate the damping ratio to apply 3% for seismic analyses of Reactor Coolant Loop (RCL) with two-point-support Steam Generators (SGs) which was normally 0.5% or 1% in the past Applications. For the revalidation, vibration tests of SGs were carried out at Unit 2 and Unit 3 in Mihama Power Station of the Kansai Electric Power Co., Inc. In the test at Mihama Unit 2, SG top was hit horizontally by the pendulum type hammering device. As a result, in the hot leg (HL) direction, 9% damping ratio has been obtained. In the test at Mihama Unit 3, electro-hydraulic actuators were installed at the top of reinforced concrete wall surrounding SG and SG upper manhole was excited. In the excitation test, frequency response curves were obtained by changing the frequency stepwise in sinusoidal wave at constant amplitude. The damping ratio has been confirmed as more than 3%, specified in JEAG 4601-1991 as standard value, in the HL perpendicular direction which provided smaller damping ratio compared to the HL direction. Dissipation energy of snubber was measured and it has been confirmed that snubbers themselves do not contribute damping effect for small SG displacement like tests in Mihama Unit 2 and Unit 3. Large dissipation energy of snubbers would be expected in earthquake. It has been realized that conservative large responses are computed in RCL seismic analysis if the damping ratios obtained are used.

Steam Generator Modal Analysis of Nuclear Power Plant

A method for forming a simplified model of steam generator which will be used in reactor coolant loop analysis has been shown here, as well as the modal analysis to this simplified SG model. This modal analysis results and the results of the SG provided by NPP designer are compared together in order to prove the design correctness. The comparison shows that the two are basically consistent.

Application of Laser Measurement and Modeling Technology in Nuclear Power Project RCL Installation

Briefly introduce reactor coolant loop (RCL) laser measurement and modeling technology. Due to the specific characteristics of nuclear power project, there is no regulating section in the RCL. It is very important to determine the position of RCL’s cutting line, and any miscalculation of cutting line may result in the RCL’s scrapping. Conduct measurement and modeling to reactor pressure vessel, steam generator and RCL by using laser measurement and 3D modeling technology; within one coordinate system, finding out cutting lines and implementing virtual cutting grouping assembly in advance through computer simulation can avoid cost increasing and construction delay caused due to wrong RCL cutting. Positioning by making use of SG (Steam Generator)-end cutting lines of RCL installation process of the nuclear power plant which is measured and modeled by laser, and the data has been implemented after being approved by supervision agency. At present, its RCL installation process has been successfully installed and practiced.

Analysis of Flashing and Slow Descending Phenomenon in Ring Coelom of Integrated Reactor During LOCA

Integrated reactor has an integrated pressure vessel with all components of the main coolant loop inside. Ring coelom is formed between the pressure vessel and the heat exchanger. Several small-bore pipes are plugged into the ring coelom to attain the coolant. Because of the existence of the isolated ring coelom, slow descending phenomenon of liquid level in the ring coelom takes place during LOCA with a coolant diversion pipe breaking. The phenomenon is analyzed by modelling the reactor during LOCA in Relap5 and one mitigation measure is acquired. As drastic flashing seriously influences the slow descending phenomenon, which enhance the pressure in the ring coelom, we divide the ring coelom into several parts and use Fluent to gain the flashing details. The results of Relap5 and Fluent show good agreement, which proves the flashing and slow descending phenomenon in the ring coelom of integrated reactor is reasonable during LOCA.