Development of a Quality Manual for Radioisotope Production Plants

The purpose of issuing a quality manual for the radioisotope production plants is to define and describe the quality system implemented by the plant. It provides general procedures for all activities comprising the quality system, defines the authorities and responsibilities of all personnel affected by the system and provides a way to inform our customers of the specific controls that are in place at radioisotope production plants to assure continued product quality. Such a quality manual is regularly updated to depict the quality management system implemented by radioisotope production plant as accurately as possible. Example of this type of quality manual is developed for the radioisotope production facility (RPF) located in ETRR-2 Complex site – Egyptian Atomic Energy Authority – Inshas-EGYPT.

A Study of Heat Transfer Calculation Method of Reactor Vessel Metallic Insulation

Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter “view factor”, and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster.

Concept of Standard System Based on China Current Nuclear Power Construction

Since introduce different technical routes, during decades of nuclear power development in our country, the French RCC series standards, American ASME standards and Russian standards are adopted, which led to the current various standards exist in their own way. To promote the building of nuclear power standards system in China, in the year of 2012, important research subject “the research on the standard system of advanced nuclear power in China” has been carried out and subject “nuclear power construction and commissioning” is one of it.. By digestion and absorption of four oversea AP1000 units of Sanmen nuclear power plant in Zhejiang province and Haiyang nuclear power plant in Shandong province, the building of standard system during nuclear power construction suitable to our national condition is studied, including the system frame and composition standards, building standard system method during construction, namely through research and example to present what kind of standard system is suitable for China standard system during construction, and what kind of method or design is used to obtain and maintain such system. The thesis is to promote the subject research methods based on examples to build China’s nuclear power standard system.

Global Balance of Plant in NPPs Using Process Data Reconciliation According to VDI 2048

The large quantities of measurement information gathered throughout a plant process make the closing of the mass and energy balance nearly impossible without the help of additional tools. For this reason, a variety of plant monitoring tools for closing plant balances was developed. A major problem with the current tools lies in the non-consideration of redundant measurements which are available throughout the entire plant process. The online monitoring reconciliation system is based on the process data reconciliation according to VDI 2048 standard and is using all redundant measurements within the process to close mass and energy balances. As a result, the most realistic process with the lowest uncertainty can be monitored. This system is installed in more than 35 NPPs worldwide and is used ○ as a basis for correction of feed water mass flow and feed water temperature measurements (recover of lost Megawatts). ○ as a basis for correction of Taverage (Tav) (recover of steam generator outlet pressure in PWRs). ○ for maintaining the thermal core power and the feed water mass flow under continuous operation conditions. ○ for automatic detection of erroneous measurements and measurement drift. ○ for detection of inner leakages, non-condensable gases and system losses. ○ for calculating non measured values (e.g. heat transfer coefficients, ΔT, preheater loads,…). ○ as a monitoring system for the main thermodynamic process. ○ for verifying warranty tests more accurate. ○ as a application of condition-based maintenance and component monitoring. ○ for What-if scenarios (simulation, not PDR) This paper describes the methodology according to VDI 2048 (use of Gaussian correction principle and quality criterias). The benefits gained from the use of the online monitoring system are demonstrated.

Evaluation of Fatigue Life of Pressurized Water Reactor Internals Considering Light-Water Reactor Coolant Environmental Effect for Aging Management Program

Most of the current operating nuclear power plants in the United States were designed using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for fatigue design curves. These design curves were developed in the late 1960s and early 1970s. They were often referred to as “air curves” because they were based on tests conducted in laboratory air environments at ambient temperatures. In recent years, laboratory fatigue test data showed that the light-water reactor environment could have significant impact on the fatigue life of carbon and low-alloy steels, austenitic stainless steel, and nickel-chromium-iron (Ni-Cr-Fe) alloys. United States Nuclear Regulatory Commission, Regulatory Guide 1.207 provides a guideline for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors. It recommend following the method developed in NUREG/CR-6909 [3] when designing reactor coolant pressure boundary components. The industry has invested a lot of effort in developing methods and rules for applying environmental fatigue evaluations for ASME Class 1 components and piping. However, the industry experience in applying the environmental fatigue evaluation for reactor core support structures and internal structures has been very limited. During the recent aging management programs, reactor internal component environmental fatigue evaluations for several pressurized water reactors were evaluated. The analyses calculated the cumulative fatigue usage using the recorded plant-specific transient cycles and the projected cycles for 60 years of plant life. The study concludes that the actual fatigue usages of the components are substantially lower than the specified original design conditions. Even assuming the most severe light-water reactor coolant environmental effects, fatigue will not be a concern for 60 years of plant life. The experiences with environmental fatigue evaluation for reactor internals are still very limited. This study shall provide the industry with beneficial information to develop the approaches and rules addressing the environmental effect on the fatigue life of reactor internals.

The Use of Experimental Design for the Shrink-Fit Assembly of Multi-Ring Flywheel

The flywheel of latest coolant pump provides high inertia to ensure a slow decrease in coolant flow to prevent fuel damage after the loss of power. Flywheel comprises a hub, twelve tungsten alloy blocks and a retainer ring shrink-fit assembled on the outer surface of blocks. In the structural integrity analysis, the shrinkage load due to shrink-fit and the centrifugal load due to rotation are considered, so the wall thickness of retainer ring and the magnitude of shrink-fit are key variables. In particular, these variables will change the flywheel running state. This paper considers the influence of these variables, we employ Latin hypercube design to obtain the response surface model and analyze the influence of these variables. Finally we obtain the magnitude of wall thickness of retainer ring and the range of shrink-fit.

Loss of Representativeness of the Fixed Incore Instrumentation of the EPR™ Reactor

The EPR™ reactor features a fixed incore instrumentation, composed of 72 Self Powered Neutron Detectors (SPND), that provides the online reconstruction of the core maximum Linear Power Density (LPD) and minimum Departure from Nucleate Boiling Ratio (DNBR). The Instrumentation and Control (I&C) systems of the EPR™ reactor use this online reconstruction in surveillance and protection functions. The onsite thresholds of those I&C functions have to take into account all the uncertainties affecting the online reconstruction of core power distribution measured by SPNDs. One of these uncertainties is the so-called Loss Of Representativeness (LOR). This uncertainty is defined as the difference between the LPD (respectively DNBR) physical value and the LPD (respectively DNBR) computed value using SPND signals. The LOR parameter is mostly linked to the difference between the core power distribution at the time where SPNDs are calibrated and the core power distribution at the time where their signals are used. For the DNBR, LOR also takes into account the use of a simplified on-line DNBR calculation algorithm. A statistical approach is used in order to define this uncertainty. The analysis is based on the evaluation of different sets of core power distributions generated thanks to random drawings of the plant state parameters (including power level, core inlet temperature, pressure, control rod insertion and xenon distribution). The sets of core configurations representative of normal plant operation are used to define the surveillance thresholds. The sets representative of accidental transients (for which the LPD and DNBR protections are claimed) are used to define the protection thresholds. The analysis of LOR values provides an envelop probability law covering a minimum of 95% of LOR values. In order to derive the on-site threshold for LPD and DNBR, a Monte Carlo method is used to propagate the LOR probability law and the other uncertainties. Sensitivity calculations have been performed in order to cover a large spectrum of fuel loading patterns and to take into account SPND failures. In conclusion, this approach allows defining an optimized and robust set of thresholds for the on-line surveillance and protection system of EPR™ reactor.

An Experimental Study on Insulating Property of Helium in High Temperature Gas Cooled Reactor

HTGR, short for high temperature gas cooled reactor, has gained a lot of attention in nuclear industry. Gas helium, 7MPa in pressure, is used as primary coolant of HTR-PM in where there are a lot of electrical equipment. Insulating property of helium is worse than that of air according to Paschen curves and there are very few articles or related standards about insulating property of high pressure gas helium, which makes the electrical equipment structure design lack of basis. In this study, an experimental platform for testing insulating performance is designed, based on which the experiments of testing the withstanding voltages of penetration assemblies and the breakdown voltages of parallel plane electrodes at different pressures are carried out. The results show that for both the penetration assemblies and the parallel plane, their breakdown voltages in helium are far lower than in air under the same condition of 15°C /0.1MPa. For the penetration assemblies, their insulating properties in helium at 150°C/7MPa are better than those in air at 15°C/0.1MPa.

Testing of Statistical Procedures for Use in Optimization of Reactor Performance Under Aged Conditions

This paper discusses a framework for designing artificial test problems, evaluation criteria, and two of the benchmark tests developed under a research project initiated by the Canadian Nuclear Safety Commission to investigate the approaches for qualification of tolerance limit methods and algorithms proposed for application in optimization of CANDU reactor protection trip setpoints for aged conditions. A significant component of this investigation has been the development of a series of benchmark problems of gradually increased complexity, from simple “theoretical” problems up to complex problems closer to the real application. The first benchmark problem discussed in this paper is a simplified scalar problem which does not involve extremal, maximum or minimum, operations, typically encountered in the real applications. The second benchmark is a high dimensional, but still simple, problem for statistical inference of maximum channel power during normal operation. Bayesian algorithms have been developed for each benchmark problem to provide an independent way of constructing tolerance limits from the same data and allow assessing how well different methods make use of those data and, depending on the type of application, evaluating what the level of “conservatism” is. The Bayesian method is not, however, used as a reference method, or “gold” standard, but simply as an independent review method. The approach and the tests developed can be used as a starting point for developing a generic suite (generic in the sense of potentially applying whatever the proposed statistical method) of empirical studies, with clear criteria for passing those tests. Some lessons learned, in particular concerning the need to assure the completeness of the description of the application and the role of completeness of input information, are also discussed. It is concluded that a formal process, which should include extended and detailed benchmark tests, but targeted to the context of the particular application and aimed at identifying the domain of validity of the proposed tolerance limit method and algorithm, is needed and might provide the necessary confidence in the proposed statistical procedure.