Scope

From the educational point of view, there are many textbooks on reactor physics used at various universities in the world. However, most of these textbooks focus either on application to Light Water Reactors (LWRs), or on the theory and mathematics, with a significant number of equations and computational schemes. Or else they were written more than 20, or even more than 60, years ago, and therefore they do not reflect the evolution of reactor concepts and engineering requirements since then. All those categories of books are either difficult to follow for non-physicists working in the nuclear industry, or else are of little value for those who are interested in special features of CANDU reactor physics.

Long.Term Reactivity Change and Control: On.Power Refuelling

There are 2 concepts related to the “age” of fuel: irradiation (fluence) and fuel burnup. The fuel irradiation in a given fuel bundle, denoted ω, is defined as the time integral of the thermal flux in the fuel during its residence time in the core. Another term for irradiation is fluence. Irradiation is also known as the thermal-neutron exposure of the fuel. The units of irradiation are neutrons/cm2, or more conveniently, neutrons per kilobarn, n/kb. Since the cut-off of the thermal-energy range may be defined differently in different computer codes, the fuel irradiation may vary from computer code to computer code, and caution must therefore be exercised when comparing irradiation values using different codes. In documents, it has been more and more usual to report values of fuel burnup rather than fuel irradiation, as burnup does not suffer from differences in definition between codes.

Neutron Kinetics and Reactor Control

A nuclear reactor is designed to achieve the very delicate balance between neutron “production” (release) in fission reactions and neutron loss by absorption and leakage. A given neutron will be “born” in a fission event and will then usually scatter about the reactor until it meets its eventual “death” either by being absorbed in some material or by leaking out of the reactor. A certain number of these neutrons will be absorbed by fissionable nuclei and induce further fissions, thereby leading to the birth of new fission neutrons, that is, to a new generation of neutrons. The ratio of the number of neutrons born in a fission-neutron generation to the number born in the previous generation is called the effective reactor multiplication factor, keff. The keff characterizes the balance or imbalance in the chain reaction. Alternatively, keff can be defined by the ratio of production rate to loss rate of neutrons in the reactor. These definitions are given below:

CANDU Reactor.Physics Analysis Methods and Computer Codes

Reactor physics aims to understand accurately the reactivity and the distribution of all the reaction rates (most importantly of the power), and their rate of change in time, for any reactor configuration. To do this, the multiplication factor (or, equivalently, reactivity) and the neutron-flux distribution under various operating conditions and at different times need to be calculated repeatedly. Most of the other parameters of interest (such as neutron reaction rates, power, heat deposition, etc.) are derived from them. They are governed by the geometry, the material composition and the nuclear data (i.e., the neutron cross sections, their energy dependence, the energy spectra and the angular distributions of secondary particles, etc.). For radiation-shielding calculations, additional photon interactions and coupled neutron-photon interaction data are required.

Back Matter

This back matter contains the References and Bibliography.

Short.Term Reactivity & Power Control: Reactor Regulating System (RRS)

For CANDU reactors, the control of the long-term reactivity and of the power is carried out by on-power refuelling, while the control of the short-term reactivity and of power is done by the RRS. The RRS is part of the overall plant-control system that maintains the reactor power at a specified level, or, when required, manoeuvres the reactor power between specified setpoints. The reactor power setpoint can be entered by the operator (in the reactor-leading mode) or it can be calculated automatically by the Steam Generator (SG) pressure-control program (in the turbine-leading mode). The RRS consists of the following main components:

Approach to Criticality

Source neutrons are essential for reactor restart after a long shutdown. The term “source neutrons” applied to a particular time interval refers to a steady supply of neutrons, constant over the time interval of interest. This supply must be independent of the current or very recent fission rate, which can vary over the time interval. Thus, source neutrons exclude prompt neutrons and even delayed neutrons which originate in the fuel (i.e., those born in the fuel itself). This exclusion does not apply to delayed photoneutrons, which come from fissions that have occurred a long time before, and whose numbers are quite constant over the current time interval (further discussion of this point below).

Reactor Shutdown and Reactor Restart

The power referred to most frequently in reactor physics is neutron power. Neutron power is essentially the fission rate multiplied by the average prompt energy released and recovered per fission (see Section 2.1.2). It is also called “prompt” power, as it appears very quickly following fission. We cannot measure neutron power directly, but we do monitor the neutron flux with ion chambers located outside the calandria and in-core flux detectors. These neutronic signals are calibrated to the thermal-power measurement which allows neutron power to be derived.

Basics of CANDU Reactor Physics

Nuclear fission is the splitting of a (large) nucleus, with the release of energy. The nuclei of some heavy elements, such as U-238, do exhibit spontaneous fission in nature. However, the rate of such fissions is extremely low. The half-life of uranium is longer than 100 million years, and most of its decay is by alpha emission, so spontaneous fission is not a practical source of energy. Spontaneous fission is not of much use to us as an energy source!

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.