Thư viện tri thức trực tuyến
Kho tài liệu với 50,000+ tài liệu học thuật
© 2023 Siêu thị PDF - Kho tài liệu học thuật hàng đầu Việt Nam
Tài liệu THERMAL-HYDRAULIC IN NUCLEAR REACTOR pptx
Nội dung xem thử
Mô tả chi tiết
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
GS. Trần Đại Phúc
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
Summary
1.Introduction
2.Energy from fission
3.Fission yield
4.Decay heat
5.Spatial distribution of heat sources
6.Coolant flow & heat transfer in fuel rod assembly
7.Enthalpy distribution in heated channel
8.Temperature distribution in channel in single phase
9.Heat conduction in fuel assembly
10.Axial temperature distribution in fuel rod
11.Void fraction in fuel rod channel
12.Heat transfer to coolant
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
I. Introduction
An important aspect of nuclear reactor core analysis
involves the determination of the optimal coolant flow
distribution and pressure drop across the reactor core. On
the one hand, higher coolant flow rates will lead to better
heat transfer coefficients and higher Critical Heat Flux
(CHF) limits. On the other hand, higher flows rates will also
in large pressure drops across the reactor core, hence
larger required pumping powers and larger dynamic loads
on the core components. Thus, the role of the
hydrodynamic and thermal-hydraulic analysis is to find
proper operating conditions that assure both safe and
economical operation of the nuclear power plant.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
This chapter presents methods to determine the distribution
of heat sources and temperatures in various components of
nuclear reactor. In safety analyses of nuclear power plants
the amount of heat generated in the reactor core must be
known in order to be able to calculate the temperature
distributions and thus, to determine the safety margins. Such
analyses have to be performed for all imaginable conditions,
including operation conditions, reactor startup and shutdown,
as well as for removal of the decay heat after reactor
shutdown. The first section presents the methods to predict
the heat sources in nuclear reactors at various conditions. The
following sections discuss the prediction of such parameters
as coolant enthalpy, fuel element temperature, void fraction,
pressure drop and the occurrence of the Critical Heat Flux
(CHF) in nuclear fuel assemblies
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
I.1. Safety Functions & Requirements
The safety functions guaranteed by the thermal-hydraulic
design are following:
Evacuation via coolant fluid the heat generated by the
nuclear fuel;
Containment of radioactive products (actinides and fission
products) inside the containment barrier.
Control of the reactivity of the reactor core: no effect on the
thermal-hydraulic design.
Evacuation of the heat generated by the nuclear fuel: The
aim of thermal-hydraulic design is to guarantee the
evacuation of the heat generated in the reactor core by the
energy transfer between the fuel
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
Rods to the coolant fluid in normal operation and incidental
conditions.
The thermal-hydraulic design is not under specific design
requirements.
However, the assured safety functions requires the
application of a Quality Assurance programme on which the
main aim is to document and to control all associated
activities.
Preliminary tests: The basic hypothesis on scenarios
adopted in the safety analyses must be control during the
first physic tests of the reactor core. Some of those tests,
for example the measurements of the primary coolant rate
or the drop time of the control clusters, are performed
regularly. Other tests are performed in totality only on the
head of the train serial.
For the following units, only the necessary tests performed
to guarantee that thermal-hydraulic characteristics of the
reactor core are identical to the ones of the head train
serial.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
The primary coolant rate and the drop time of the control rod clusters
must be measured regularly.
The main aim of the thermal-hydraulic design is principally to
guarantee the heat transfer and the repartition of the heat production
in the reactor core, such as the evacuation of the primary heat or of
the safety injection system (belong to each case) assures the respect
of safety criteria.
I.2. Basis of thermal-hydraulic core analysis
The energy released in the reactor core by fission of enriched uranium
U235 and Plutonium 238 appears as kinetic energy of fission reaction
products and finally as heat generated in the nuclear fuel elements.
This heat must be removed from the fuel and reactor and used via
auxiliary systems to convert steam-energy to produce electrical
power.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
I.3. Constraints of the thermal-hydraulic core design
The main aims of the core design are subject to several
important constraints.
The first important constraint is that the core temperatures
remain below the melting points of materials used in the
reactor core. This is particular important for the nuclear
fuel and the nuclear fuel rods cladding.
There are also limits on heat transfer are between the fuel
elements and coolant, since if this heat transfer rate
becomes too large, critical heat flux may be approached
leading to boiling transition. This, in turn, will result in a
rapid increase of the clad temperature of the fuel rod.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
The coolant pressure drop across the core must be kept low to
minimize pumping requirements as well as hydraulic loads
(vibrations) to core components.
Above mentioned constraints must be analyzed over the core live,
for all the reactor core components, since as the power
distribution in the reactor changes due to fuel burn-up or core
management, the temperature distribution will similarly change.
Furthermore, since the cross sections governing the neutron
physics of the reactor core are strongly temperature and density
dependent, there will be a strong coupling between thermalhydraulic and neutron behaviour of the reactor core.
II. Energy from nuclear fission
Consider a mono-energetic neutron beam in which n is the
neutron density (number of neutrons per m3). If v is neutron
speed then Snv is the number of neutron falling on 1 m2 of target
material per second.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
Since s is the effective area per single nucleus, for a given
reaction and neutron energy, then S is the effective area of
all the nuclei per m3 of target. Hence the product Snv gives
the number of interactions of nuclei and neutrons per m3 of
target material per second.
In particular, the fission rate is found as: Σf nv = ΣfФ ,
where Σf =nv is the neutron flux (to be discussed later) and
Σf= Nσf
, N being the number of fissile nuclei/m3 and σf
m2/nucleus the fission cross section. In a reactor the
neutrons are not mono-energetic and cover a wide range of
energies, with different flux and corresponding cross
section.
In thermal reactor with volume V there will occur V Σf Ф
fissions, where Σf and Ф are the average values of the
macroscopic fissions cross section and the neutron flux,
respectively.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
To evaluate the reactor power it is necessary to know the
average amount of energy which is released in a single
fission. The table below shows typical values for uranium235.
Table II.1: Distribution of energy per fission of U-235.
10-12 J = 1 MeV
Kinetic energy of fission products 26.9 168
Instantaneous gamma-ray energy 1.1 7
Kinetic energy of fission neutrons 0.8 5
Beta particles from fission products 1.1 7
Gamma rays from fission products 1.0 6
Neutrinos 1.6 10
Total fission energy 32 200
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
As can be seen, the total fission energy is equal to 32 pJ. It
means that it is required ~3.1 1010 fissions per second to
generate 1 W of the thermal power. Thus, the thermal
power of a reactor can be evaluated as:
P (W) = VΣfФ / 3.1x1010 (W)
Thus, the thermal power of a nuclear reactor is
proportional to the number of fissile nuclei, N, and the
neutron flux f . Both these parameters vary in a nuclear
reactor and their correct computation is necessary to be
able to accurately calculate the reactor power.
Power density (which is the total power divided by the
volume) in nuclear reactors is much higher than in
conventional power plants. Its typical value for PWRs is 75
MW/m3, whereas for a fast breeder reactor cooled with
sodium it can be as high as 530 MW/m3.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
III. Fission yield
Fissions of uranium-235 nucleus can end up with 80
different primary fission products. The range of mass
numbers of products is from 72 (isotope of zinc) to 161
(possibly an isotope of terbium). The yields of the products
of thermal fission of uranium-233, uranium-235,
plutonium-239 and a mixture of uranium and plutonium are
shown in following figure III.1.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
Figure III.1: Fission yield as a function of mass number of
the fission product.
As can be seen in all cases there are two groups of fission
products: a “light” group with mass number between 80
and 110 and a “heavy” group with mass numbers between
125 and 155.
THERMAL-HYDRAULIC IN
NUCLEAR REACTOR
Figure III.2: Illustration of the 6 formula: