Regional/Neutron Overpower Protection (ROP/NOP)

There are independent, separate, and diverse ROP/NOP systems for the two SDS. Each ROP/NOP system consists of a number of flux detectors (see Section 5.4.3) which provide prompt measurements of neutron flux throughout the core. The detectors are mounted inside assemblies that penetrate the core, perpendicular to the fuel channels, vertically or horizontally. The system for SDS1 uses vertical detectors, the one for SDS2 uses horizontal detectors. Detectors are judiciously distributed to monitor the neutron flux throughout the core. In the CANDU 6, a total of 58 ROP detectors are used: 34 for SDS1 and 24 for SDS2. The number and location of detectors in the core are selected in an analysis whose objective is to ensure that as small a number of detectors as possible protect the reactor by tripping a SDS when local high powers threaten reactor safety from any flux shape that could arise in the operating reactor, while at the same time providing adequate margin-to-trip (MTT) to avoid possible restrictions on reactor operating power.

Short.Term Reactivity Change: Local.Parameter Reactivity Feedback

A change in power for an operating reactor generally alters local parameters in the reactor such as the temperatures of the fuel, moderator, and coolant. A change in any of these local parameters causes a change in reactivity that, in turn, affects reactor operation (a feedback effect). Local parameters help to understand the feedback reactivity components related to the core evolution. For the CANDU reactor, the most important local parameters are the following:

Biogas Production from Organic Wastes: Integrating Concepts of Circular Economy

Anaerobic digestion is traditionally used for treating organic materials. This allows the valorization of biogas and recycling of nutrients thanks to the land application of digestates. However, although this technology offers a multitude of advantages, it is still far from playing a relevant role in the energy market and from having significant participation in decarbonizing the economy. Biogas can be submitted to upgrading processes to reach methane content close to that of natural gas and therefore be compatible with many of its industrial applications. However, the high installation and operating costs of these treatment plants are the main constraints for the application of this technology in many countries. There is an urgent need of increasing reactor productivity, biogas yields, and operating at greater throughput without compromising digestion stability. Working at organic solid contents greater than 20% and enhancing hydrolysis and biogas yields to allow retention times to be around 15 days would lead to a significant decrease in reactor volume and therefore in initial capital investments. Anaerobic digestion should be considered as one of the key components in a new economy model characterized by an increase in the degree of circularity. The present manuscript reviews the digestion process analyzing the main parameters associated with digestion performance. The novelty of this manuscript is based on the link established between operating reactor conditions, optimizing treatment capacity, and reducing operating costs that would lead to unlocking the potential of biogas to promote bioenergy production, sustainable agronomic practices, and the integration of this technology into the energy grid.

OSCAR-4 CODE SYSTEM COMPARISON AND ANALYSIS WITH A FIRST-ORDER SEMI-EMPIRICAL METHOD FOR CORE-FOLLOW DEPLETION CALCULATION IN MNR

Knowledge of the isotopic composition of a nuclear reactor core is important for accurate core-follow and reload analysis. In the McMaster Nuclear Reactor, fuel depletion estimates are based upon a semi-empirical calculation using flux-wire measurements. These estimates are used to plan and guide fuelling operations. To further support operations, an OSCAR-4 model is being developed. To evaluate the performance of the OSCAR-4 code for this application, 2 points of comparison, considering the period between 2007 and 2010, are presented: (i) the multiplication factor keff and (ii) U-235 fuel inventory. The latter is compared with a simple first-order semi-empirical calculation. The calculation of keff for the last operational 3 months yields 0.997 ± 0.002 (vs. 1.000 for an operating reactor), and differences in both core-average inventory and the maximum standard fuel assembly inventories estimates are found to be 5.7% and 7.5%, respectively.

PREDICTION OF A NEUTRON FLUENCE BY AN ULTRASONIC MEASUREMENT IN AN IRRADIATED MATERIAL USED FOR SURVEILLANCE TESTING OF AN OPERATING REACTOR

The ultrasonic attenuation versus neutron fluence relationship from Charpy impact test specimens withdrawn from an operating reactor vessel during plant outages has been evaluated in order to predict the neutron fluence through a direct attenuation measurement for a reactor vessel material irradiated by neutrons. The results showed that systematic relationships for the ultrasonic attenuation versus neutron fluence at specific operating frequencies were observed for the base and weld metal and the possibility to predict neutron fluence in a specimen, that has no information on the amount of neutron fluence, by using such a consistent relationship.

In situ study of the 41Ar plume released from the Ignalina NPP

The 41Ar gamma ray radiation was registered using the in situ method in the vicinity of the Ignalina nuclear power plant (NPP). The sum of gamma rays, that are reaching the HPGe detector, situated along the wind direction, from a number of plume segments and which are registered in the energy range of about 1.29 MeV, is calculated. An independent technological regime of the operating reactor method of the determination of the 41Ar emission rate from NPP stack is introduced.

Insights and Observations Arising From Curve-Fitting the Charpy V-Notch and Tensile Data Contained Within the United States’ Light Water Reactor Surveillance Database

The effect of irradiation damage on the mechanical properties of reactor pressure vessel structures is monitored in operating nuclear reactors according to the provisions of 10 CFR Part 50 Appendix H. In these surveillance programs Charpy V-notch energy and tensile data are collected. Trends in these data have and continue to be used to identify and quantify embrittlement trends, which is a key aspect to maintaining the continued structural integrity of the operating reactor fleet. This paper presents the current results from an on-going investigation aimed at assessing the effect of different curve-fitting strategies on the insights that can be gained from these data trending activities.