report on failure modes in reactor pressure vessel -incomplete

8/10/2019 Report on Failure Modes in Reactor Pressure Vessel -Incomplete

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Report on possible hazards due to failure of the

reactor pressure vessel of a pressurized waterreactor

Thesis submitted in the fulfilment of the requirements for the award of the

degree of

MASTER OF TECHNOLOGY (HONS.)

IN

INDUSTRIAL ENGINEERING

BY

B.Venkata Sainath

Roll No: 10MF3IM05

Under the guidance of

Prof. J.Maiti

DEPARTMENT OF INDUSTRIAL & SYSTEMS

ENGINEEERING

INDIAN INSTITUTE OF TECHNOLOGY KHARAGPUR


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Table of Contents

1. Introduction:-...................................................................................................................................... 4

2. Reactor pressure vessel:-..................................................................................................................... 5

2.1. Reactor vessel body:-................................................................................................................... 5

2.2. Reactor core:-............................................................................................................................... 7

2.3. Coolant/moderator:-................................................................................................................... 11

3. Failure Mode and Effect Analysis (FMEA) of Reactor Pressure vessel of PWR........................... 14

3.1 Reactor pressure vessel (System ) breakdown using hardware approach (bottom-up

approach)........................................................................................................................................... 14

3.2. Failure modes of the components of Reactor pressure vessel .................................................... 15


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Table of figures

Figure 1Cutaway View of Reactor Vessel ................................................................................................ 6

Figure 2 PWR fuel assembly with control rods ....................................................................................... 8

Figure 3 Nuclear fission reaction of Uranium-235 ............................................................................... 12

Figure 4 Breakdown of reactor pressure vessel into components ....................................................... 14
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1. Introduction:-

Since the demonstration of a sustained fission reactor in 1942, nuclear

power has emerged as a proven technology and as a method for producing

electricity in the world. Because of the worlds continuously improving living

standards, increased population and concern over the increased concentration of

greenhouse gas emissions caused by burning fossil fuels, it is not surprising

that there is likely to be an increasing demand for nuclear power.

About 5.7% of the worlds energy and 13% of the worlds electricity in

2012 was generated from the nuclear power plants (excluding the contribution

from naval nuclear fission reactors)[1]

.Since power plants are very complex

systems ,consisting of many sub-systems and thousands of components,

operation and maintenance of them is a challenging task. Sometimes, failure of

a single component may lead to the failure of the subsystem which can

drastically affect the operation of the plant. One of the most important

components in a nuclear power plant is the reactor pressure vessel where the

nuclear fission reaction takes place and heat is generated. The present study

aims at identification of different failure modes of the reactor pressure vessel

and possible hazards due to its failure.


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Figure 1Cutaway View of Reactor Vessel


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The core barrel slides down inside of the reactor vessel and contains the

reactor core assembly. There exists a lower support plate towards the bottom of

the core barrel on which fuel assemblies sit. The core barrel and all of the lower

internals actually hang inside the reactor vessel from the internals support ledge.

Neutron shield panels are attached to the core barrel opposite the core corners,where neutron flux tends to be higher. There will be irradiation specimen

holders on the outside of the core barrel in which samples of the material used

to manufacture the vessel will be placed. These samples are removed and tested

at periodic intervals to examine the influence of radiation on the strength of the

material.

2.2. Reactor core:-It is the heart of the nuclear power plant since it contains the nuclear fuel

components where the nuclear reaction takes place. It is the region where the

nuclear fuel assemblies are located and all of the heat is generated in a nuclear

reactor. It is one of the most complicated systems in the nuclear power plant and

consists of hundreds of fuel assemblies, control rods, instrumentation guide

tubes, sensor for measuring coolant levels, core supporting components likecore shroud, core support column etc,. But it can be divided into four sub-

systems for based on the functionality as follows:

1. Fuel rod assembly

2. Control rod assembly

3. Core support assembly

4. Shim rods (burnable poisons)


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1 Fuel rod assembly:

The typical nuclear fuel used in the nuclear reactor core of a nuclear power

plant is cylindrical pellets of uranium dioxide(An alternate pellet used is a mix

of natural uranium and reactor grade plutonium,240

Pu; this is called mixedoxide fuel, or MOX).

Slightly enriched UO2 pellets (3.3%) of diameter about 9mm are stacked

inside a zirconium alloy tube of length 3.8m and thickness 0.64mm known as

fuel rod or fuel pin. These fuel rods are assembled in a 17*17 square pattern pin

locations to form a fuel assembly. However, in many of these assemblies, 24

locations are occupied by guide tubes in which control-rod "fingers," held at the

top by a "spider," move up and down in the assembly to provide coarse

reactivity control. Fuel assemblies of around 200-300 are loaded vertically in a

nuclear reactor core. Generally, fuel assemblies spend around 3 years before

they are replaced by the new ones.

Figure 2 PWR fuel assembly with control rods


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2. Control rod assembly

Control rodsare used in nuclear reactorsto control the fission rate

of uraniumandplutonium. They are made of chemical elements, such asboron

, silver, indiumand cadmium , that are capable of absorbing many neutronswithout fissioning themselves. Because these elements have different capture

cross sectionsfor neutrons of varying energies, the composition of the control

rods must be designed for the neutron spectrum of the reactor they are supposed

to control.

Short term or emergency reactivity control is provided by the 24 control

rod fingers in many of the assemblies. These control fingers usually contain

B^C or, more recently, a mixture of silver (80%), indium (15%), and cadmium

(5%) to produce slightly weaker absorbers. Generally, 4-9 adjacent control-rod

spiders (which connect all the control rod fingers in an assembly) are grouped

together and moved together as a single control-rod bank. The various control

rod banks then provide coarse reactivity control.

3. Core support assembly:

It includes all the components that support the reactor core structurallyand functionally. These include core shroud, instrumentation guide tubes, core

baffle, core support column, lower core plate etc.

The reactor internals consist of the lower core support structure, the upper

core support structure, and the in core instrumentation support structure. The

internals are designed to support, align, and guide the core components; direct

coolant flow to and from core components; and guide and support the in core

instrumentation. The core barrel supports and contains the fuel assemblies and

directs the coolant flow. The core barrel is a cylindrical shell 147.25 inches in

diameter and 330.75 inches long. The barrel hangs from a ledge on the reactor

vessel flange and is aligned by four flat sided pins which are press fitted into the

barrel at 90-degree intervals.

The upper core support structure provides structural support for the fuel

assemblies and in core instrumentation. The structure consists of an upper

support assembly, upper support columns, RCCA guide columns, thermocouple

columns, and the upper core plate.


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The in core instrumentation support structure consists of an upper system

to convey and support in core temperature monitors (thermocouples)

penetrating the vessel through the head and a lower system to convey and

support in core nuclear instrumentation flux thimbles penetrating the vessel

through the bottom.

4. Shim rods (burnable poisons) :-

For long-term reactivity control, burnable poisons are placed in some

of the lattice positions of the fuel assemblies. These shim rods, from 9 to 20 per

assembly, are stainless steel clad boro-silicate glass or Zirc-aloy clad diluted

boron in aluminium oxide pellets.


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2.3. Coolant/moderator:-All the PWR nuclear reactors in the world are light water reactors and

they use water as coolant and neutron moderator. Operating mechanism andfunctionality of the water as both the coolant and neutron moderator is

explained below.

a. Coolant:-

Primary Coolant enters the pressure vessel through an inlet nozzle on the

vessel from a pipeline connected to it, cools the elements inside the core and

exists through outlet nozzle.

The flow path for the reactor coolant through the reactor vessel would be

as follows [3]:

The coolant enters the reactor vessel at the inlet nozzle and hits against

the core barrel.

The core barrel forces the water to flow downward in the space between

the reactor vessel wall and the core barrel.

After reaching the bottom of the reactor vessel, the flow is turned upward

to pass through the fuel assemblies.

The coolant flows all around and through the fuel assemblies, removing

the heat produced by the fission process. Flow holes in the lower core

plates are sized to permit a higher coolant flow rate through the centre of

the core where power generation is greater than at the periphery.

The now hotter water enters the upper internals region, where it is routed

out to the outlet nozzle and goes on to the steam generator.

Large amount of heat is generated in the reactor core during the nuclearfission chain reaction. This heat must be captured from the core and transferred

for use in electricity generation. If coolant fails in removing the heat generated

by fission reaction, fuel elements inside the core may melt and it may lead to

severe nuclear accident known as core melt accident or nuclear meltdown.

This may be triggered by the unavailability of adequate coolant to remove the

heat from the core, commonly known as Loss of Coolant Accident (LOCA) or

fall of coolant pressure below the specification limit without any means to

restore it known as Loss of Coolant Pressure Accident.


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b. Neutron moderator:-

A neutron moderator is a medium that slows down the fast moving

electrons released from a nuclear reaction, thereby turning them into thermal

neutrons capable of sustaining nuclear chain reaction involving Uranium235.

Almost 75% of the worlds reactors use light water as neutron moderator.

Others include solid graphite (20%), heavy water (5%).

Most of the nuclear power plant reactors in the world are thermal neutron

reactors. In such reactors, a slow moving free electron is absorbed by the heavy

nucleus of Uranium and the Uranium atom becomes unstable and splits to two

products emitting two or three fast moving free neutrons and some amount of

energy. The nuclear fission reaction of235

U is shown in figure 3.

Figure 3 Nuclear fission reaction of Uranium-235


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Energy of the free neutrons would be around 2Mev.Since three free fast

neutrons are released in the fission reaction, the reaction can become a chain

reaction under controlled conditions. This results in the liberation of tremendous

amount of energy. The probability of further fission depends on the fission cross

section which in turn depends on its speed of neutron. Slow moving thermal

neutrons are much more likely to cause fission in thermal nuclear reactors

unlike fast neutrons.

The nuclear cross section of uranium-235 for slow thermal

neutrons is about 1000 barns, while for fast neutrons it is in the order of 1 barn.

Newly released fast neutrons have a velocity of around 10% of light velocity.

They must be slowed down to a speed of few km/sec for the nuclear fission

reaction to occur and for the continuation of chain reaction. This vital task is

performed by the moderator. When the fast moving neutrons collide with

nucleus of atoms of moderator, kinetic energy is transferred from fast moving

neutron to atoms of moderator. After a series of collisions, fast neutron turns

into thermal neutrons enabling sustaining the fission reaction.

In some light water nuclear reactors, boric acid is added to moderator to

absorb the neutrons. Such a soluble neutron absorber is called a chemical

shim. By varying the concentration of the boric acid, extent of absorption ofneutrons by moderator can be varied. This helps in reduction of excessive

movement of control rods.


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3. Failure Mode and Effect Analysis (FMEA) of Reactor

Pressure vessel of PWR

3.1 Reactor pressure vessel (System ) breakdown using hardware

approach (bottom-up approach)

Figure 4 Breakdown of reactor pressure vessel into components


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3.2. Failure modes of the components of Reactor pressure vessel

Total components identified = 20

I. Sub-assembly 1: Coolant

Component 1 - Coolant:-

Failure modes: 1. LOCA (Loss of Coolant Accident)

2. LCP (Loss of Coolant Pressure)

1. Loss of Coolant AccidentLOCA[2]

:-

There will be either physical loss of coolant or insufficient flow rate of

the coolant in loss of coolant accident (LOCA). In a loss of forced circulationaccident, a gas cooled reactors circulators (generally motor or steam driven

turbines) fail to circulate the gas coolant within the core. This results in

impeding of heat transfer through forced circulation though natural circulation

through convection will keep the fuel cool as long as the reactor is not

depressurized.

2. Loss of Coolant Pressure -LCP:-

In a loss of pressure control accident, the pressure of the confined coolantfalls below specification limit without any means to restore it. This results in the

reduction of heat transfer efficiency if coolant is an inert gas and in some cases,

formation of an insulating bubble of steam surrounding the fuel assemblies

(for pressurized water reactor).

Causes of LOCA: 1) physical loss of coolant

2) Insufficient flow rate of coolant

Consequences of LOCA: 1) melting of fuel rods

2) may lead to nuclear meltdown

Causes of LCP: 1) pressure of the coolant falls below specification limit

Consequences of LCP: 1)Formation of an insulating bubble of steamsurrounding fuel assemblies


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Sl.

No

Name of the

component

Location of

the

component.

Function of

the

component

Failure mode

of the

component

Cause(s) of

each failure

mode

Consequence(s)

of each failure

mode

1 Coolant Through out

the reactor

pressure

vessel

.Removes

heat

generated

from

nuclear

fission

reaction and

cools down

the reactor

core

components

LOCA,

LCP

physical

loss of

coolant,

Insufficient

flow rate of

coolant,pressure of

the coolant

falls below

specification

limit

melting of fuel

rods, may leadto nuclear

meltdown,Formation of

an insulating

bubble of

steam

surrounding

fuel assemblies

2 Fuel rod Middle

portion of

the reactor

core

Contains

the UO2pellets

Deformation

of the fuel rod

Thermal

stress

Causes trouble

in insertion of

control rod

3 Fuel assembly

alignment plate

upper ends

of the fuel

assemblies

and the

lower ends

of the

control rodguide tubes

interacts

with the

core by

positioning

the fuel

assemblies

and theguide

tubes

[6] cracking

of fuel

assembly

alignment

pins

Thermal

stress

Dislocation of

guides tubes,

inadvertent rate

of coolant flow

from lower

core part to

upper core part

4 Guide tubes

Surrounding

fuel rods

Support to

Fuel rods

Swelling of

the guide

tubes

,cracking

Thermal

stress

Leakage of

radioactive

material into

primary coolant

is high

5 Control rods Upper part

of the

reactor core

and inbetween the

fuel rods

Absorption

of neutrons

and

controllingrate of

nuclear

fission

reaction

Slower rate of

insertion,[4]Inadvertent

control rodwithdrawal,

cracking of

control rods

Deformation

of the fuel

rod

boron carbide

being dissolved

into the

primary coolant(in case of

cracking)

6 Control rod

shroud tubes Surrounding

control rods

Provides

cylinder for

hollow

piston

Distort; higher

friction

Residual

stress

Reduced drive

life

7 Control rod

spider

Upper part

of the core,below the

connect

all thecontrol rod

Distortion Deformation

of the fuelrod

Improper

insertion ofcontrol rods in


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pressure

vessel head

fingers in an

assembly)

the guide tubes

affects

reactivity

control

8 Reactor vessel

head

top of the

reactorvessel body

It contains

penetrationsto allow

the control

rod driving

mechanism

to attach to

the control

rods in the

fuel

assembly.

cracking of

the reactorvessel upper

head internals,

around some

of the control

drive

penetrations

Thermal

stress

Leakage of

coolant,LOCA,

Ejection of

control rod,

temporary

shutdown of

the reactor

9 Shim rods lattice

positions ofthe fuel

assemblies[6]

long-term

reactivitycontrol

Failure in

insertion ofShim rod

Failure of

the button

Increase in

fission rate andheat released

10 Instrumentation

Guide tubes

11 Core Baffle surround the

outer faces

of the

peripheral

fuel

assemblies

directs

coolant flowthrough thecore

cracking of

baffle former

bolts

age-related

inter

granular

stress-

corrosion

crackingprocess

influenced

by bolt

material,

fluence,

stress, and

temperature

12 Core support

column

13 Neutron shield

pad

Attached to

the corebarrel

attenuate

fastneutrons

that would

otherwise

excessively

irradiate

and

embrittle

the vessel

walls,

attenuate

gamma

radiation

Thermal

stresses areinduced in the

vessel

Embrittlement

of the reactor

pressure wall


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14 Irradiation

specimen guide

attached to

neutron

shield pad

15 Lower core

support plate

Bottom of

the core

barrel,below the

fuel

assembly

carries

the weight

of the fuelassemblies

and

distributes

the coolant

flow to the

fuel

assemblies

16 Core barrel Between the

reactor

vessel body

and reactorcore

It acts as a

supporting

structure

andcontains the

fuel

assemblies

and direct

the coolant

flow.[7]

Corrosion of

the vessel

Contamination

of the coolant

17 Inlet nozzle

Of the pressure

vessel

Entering the

coolant into

the reactor

pressurevessel

Crack-like

separations,

Failure of

nozzle weld

normal

stress at the

nozzle

Loss of coolant

pressure

18 Outlet nozzle

Of the pressure

vessel

Exiting the

coolant into

Steam

generator

from the

pressure

vessel

Crack-like

separations,

Failure of

nozzle weld

normal

stress at the

nozzle

19 Outer wall of

pressure vessel

1)Vital

safety

barrier tofission

product

release

2)Directs

reactor

coolant

Radiation

embrittlement,

fatigue

Thermal

radiation,

Residualstress

Release of

radioactive

materials toatmosphere

20 Core shroud Between

reactor core

and core

barrel

Directing

the coolant

flow ,

protecting

reactor core

Stress

corrosion

cracking[8]

Heat from

the nuclear

reactions

combined

withconstant

Nuclear

meltdown


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flowing

water

eventually

wear out the

steel plates

of coreshroud

4. Conclusion:-

Reactor pressure vessel system has been divided into sub-assemblies and

then into its components. Failure modes of each of these components and causes

and consequences of the failure mode have been identified.


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5. References:-

1.

"Nuclear Energy" - Energy Education is an interactive curriculum

supplement for secondary-school science students, funded by the U. S.Department of Energy and the Texas State Energy Conservation Office

(SECO). U. S. Department of Energy and the Texas State Energy

Conservation Office (SECO). July 2010. Retrieved 2010-07-10.

2. Source: http://en.wikipedia.org/wiki/Nuclear_meltdown

3. Reactor concepts manual on Pressurized water reactor systems by United

States Nuclear Regulatory Commission (U.S NRC).

4.Nuclear Power Plant Operating Experiences from the IAEA / NEA

Incident Reporting System 1996-1999

5.

Nuclear Regulatory commission U.S.- ABWR -RS-5146900 Rev. 1-

Design Control Document/Tier 2-15B Failure Modes and Effects

Analysis (FMEA)

6.

Source :

http://energy.gov/sites/prod/files/hss/Enforcement%20and%20Oversight/Enforcement/docs/els/Boulden_to_Grossenbacher_BEA_Sept2009.pdf

7. Westinghouse Technology Systems Manual -Section 3.1 -Reactor Vessel

and Internals

8.

http://en.wikipedia.org/wiki/Core_shroud

9. U.S.NRC - Information Notice No. 98-11: Cracking of Reactor Vessel

Internal Baffle Former Bolts in Foreign Plants-March 25, 1998

10.3.1-7 Westinghouse Technology Systems Manual -Section 3.1 -Reactor

Vessel and Internals

11.Bottom nozzle failure mechanism of water reactor pressure vessel under

severe accident conditions -Young J. Oh , Joon Lim, Kwang J. Jeong, Il

S. Hwang

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