Fire Probabilistic Risk Assessment applied to the 
Pebble Bed Modular Reactor 
Matoma P. Pooe 
2001 
l t•?H r • • 
,, . r, 
I'll" f I 
( tl "-· 
" • 1 l - ·.'1- I 9 
University 
Of North-West p B M R 
Masters in Applied Radiation Science and Technology 
University of North West, Mmabatho 
Fire Probabilistic Risk Assessment 
University p 
Of North-West B M R 
Fire Probabilistic Risk Assessment Applied to the 
Pebble Bed Modular Reactor 
Matoma P Pooe 
2 00 1 
Mr. M Magugumela &: Mr. B Blignaut 
Supervisors 
Masters in Applied Radiation Science and Technology 
University of North West , Center of Applied Radiation Science 
and Technology , Mrnabatho 
Pebble Bed Modular Reactor , Centurion 
iii 
Abstract 
A methodology for assessing the frequencies of fire occurrences 
in the PBMR nuclear power plant is presented. Fire is a concern 
because it may lead to loss of support systems, which could 
result in the failure of dependent systems. Ultimately this may 
lead to the re lease of radioactive material into the atmosphere. 
The methodology used is probabilistic in nature. The methodology 
will be applied to the Pebble Bed Modular Reactor to determine 
the impact of fire on safety of plant personnel and the public . 
The methodology described here utilises both a qualitative and 
quantitative assessment in order to quantify the risk posed by 
fires in the PBMR reactor . The qualitative steps identify areas 
where fire is l ikely to be started . Only fire events that may 
result in the loss of a safety function are of safety concern. 
Thus , this process is based on identification of critical 
locations where there is potential for fire initiation resulting 
in loss of safety function lu ~rAiv I 
The quantitative analysis utilizes an event tree approach to 
quantify the frequency of fire initiation . An event tree is 
constructed based on the mitigating actions of the fire 
protection system in each critical area . 
This study is meant to be a preliminary assessment of Fire PRA 
for the PBMR . This task is made difficult by the fact that the 
iv 
plant is still in the design phase. Traditional Fire PRA's have 
been done on existing power stations with operational histories. 
To demonstrate the methodology in the PBMR, data from Light Water 
Reactors (LWR) and Boiling Water Reactors (BWR) was used as a 
baseline for the Pebble Bed Modular Reactor (PBMR) since there is 
no historical information on the plant because one has not been 
built yet. When the plant is built and is in operation the 
assessment will need to be updated. 
A brief summary of the study, the conclusion and recommendations 
for this study are also presented in the last section of the 
document. 
V 
Table Of contents 
Abstract iii 
List of Abbreviations ... . .......... . . . . . ......... . .................. vii 
CHAPTER 1 1 
1.1 Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 
1. 2 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 
1. 3 Overview of PRA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 
1. 3 .1 Why is fire of concern . . ............................. . .. . . 3 
1 . 3.2 Historical Evidence ............ . .......... . ............... 4 
1. 3. 3 Causes of fire ............ . ..... . ... .. . ... ...... ... ....... 4 
1.4 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 
1.4.1 The Pebble Bed Modular Reactor Power plant . ............... 7 
1.5 Purpose of study ............... . . .... ..... . .. . .. ....... .. 11 
1. 6 Objectives .......... .. ..................... .... . ......... 12 
1. 7 Outline of chapters ..... .. . .. ........... . ................ 12 
CHAPTER 2 14 
2.1 Review of Literature ...................... ...... . . ....... 14 
2.2 General fire analysis methodology in power plants . ....... 25 
CHAPTER 3 27 
3.1 Fire Hazard analysis ... . ....... . . .. .... ... ..... ... ..... . . 27 
3 . 1.1 Location Screening ...... . ......... . .......... . ........... 27 
3.1.2 Fire Occurrence Frequency ....... . ..... . ... ..... .... .. .... 29 
3. 1. 3 Fire Propagation Analysis ..... .. .. . ............. . ...... . . 30 
3 .1. 4 Plant System Analysis ......... . ..... . ..... ..... .. . . ...... 30 
3.2 Application of the methodology .. . ...................... . . 32 
3.2.1 Fire Hazard Analysis ..... . ....... . .... . .......... . ...... . 32 
3.2.2 Location Screening ...... ......... . .... . ......... . ........ 40 
3.2.3 Fire Occurrence Frequency ......... . . . . . .................. 48 
3.2.4 Fire Propagation Analysis . ....... ..... ............ ... .. .. 57 
3.2.5 Plant system analysis ... .. .. . ............. . .............. 65 
CHAPTER 4 68 
4 . 1 Results ................... . .... . . .. ....... ... . ..... .. . . .. 69 
4.1.1 Fire Initiating Event Frequencies for the different rooms of 
the PBMR power plant ......................... . ..... ... ....... ... 69 
4. 2 Hypothesis ................... . .......................... . 71 
4.2.1 Fire occurrence frequencies .... . ............... . ......... 71 
4.2.2 Loss of equipment frequencies (Fire Propagation) ......... 71 
CHAPTER 5 73 
5 .1 Summary ............... . .... . . . ............ . . . ......... . .. 73 
5. 2 Conclusion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 
5. 3 Recommendations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 
Glossary of terms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 
References 81 
List of Figures 
FIGURE 1-1 GENERAL LAYOUT OF THE REACTOR UNIT AND THE POWER 
CONVERSION UNIT. 8 
FIGURE 1-2 THE PEBBLE BED MODULAR REACTOR FUEL ELEMENTS .. ... ... 11 
FIGURE 3-1 EVENT TREE FOR THE CONTROL ROOM ... .. . .. ............. 58 
FIGURE 3-2 EVENT TREE FOR THE CABLE ROOM ...... . .......... . . . ... 59 
FIGURE 3-3 EVENT TREE FOR THE ELECTRICAL SWITCHGEAR ROOM ....... 60 
FIGURE 3-4 EVENT TREE FOR THE BATTERY ROOM .... .. ...... .. . . ..... 61 
FIGURE 3-5 EVENT TREE FOR THE GENERATOR SWITCHGEAR ROOM ... ..... 62 
FIGURE 3-6 EVENT TREE FOR THE REACTOR BUILDING . .......... . ..... 63 
FIGURE 3-7 EVENT TREE FOR THE AUXILIARY BUILDING .. . . ..... . ..... 64 
List of Tables 
TABLE 3-1 SAFETY SYSTEMS AND THEIR SUPPORT SYSTEMS ............. 33 
TABLE 3-2 SAFETY SYSTEM COMPONENTS AND THEIR SUPPORT SYSTEMS 
(SYSTEM DEPENDENCIES) ............................. . . 35 
TABLE 3-3 BUILDING LAYOUT ...................................... 41 
TABLE 3-4 STATISTICAL EVIDENCE OF FIRES IN LIGHT WATER REACTORS 
AND THE RELATED FIRE FREQUENCIES .................... SO 
vi i 
LIST OF ABBREVIATIONS 
Abbreviation or acronym Description 
ACS Active Cooling System 
Arbeitsgemeinschaft Versuchsre aktor 
AVR (German for Joi ntly- operated 
Prototype Reactor ) 
BWR Boilling Water React o r 
ccs Core Conditioning System 
EPS Equipment Protection System 
FHSS Fuel Handling and Storage System 
FMEA Failure Mode and Effects Analy sis 
Failure Mode , Effects and 
FMECA 
Criticality Analysis 
HAZOP Hazards and Operability 
HICS Helium Inventory Control System 
HMS Helium Make-up System 
HPT High Pressure Turbine 
HTGR High Temperature Gas Cooled Reactor 
HTR High Temperature Reactor 
Heating, Ventilation and Air-
HVAC 
conditioning 
IAEA International Atomic Energy Agency 
IE Initiating Event 
LBE Licensing Basis Event 
LOCA Loss-of-coo l ant accident 
LOFC Loss of Forced Cooling 
LPT Low-pressure Turbine 
LWR Light Water Reactor 
MPS Main Power System 
MW Megawatt 
NNR National Nuclear Regulator 
NNS Non-nuclear Safety 
NPP Nuclear Power Plant 
NRC Nuclear Regulatory Commission (USA) 
NUREG Nuclear Regulations ( from NRC ) 
ocs Operational Control System 
PBMR Pebble Bed Modular Reactor 
PCU Power Conversion Unit 
Primary Loop Initial Clean-up System 
PES 
Evacuation Sub-system 
vii i 
PFOD Probability of Failure on Demand 
PLICS Primary Loop Initial Clean-up System 
PLOFC Pressurised Loss of Forced Cooling 
PPB Primary Pressure Boundary 
PPSB Power Plant Services Building 
PRA Probabilistic Risk Assessment 
PT Power Turbine 
PWR Pressurized Water Reactor 
PyC Pyrolytic Carbon 
RCCS Reactor Cavi ty Cooling System 
RCS Reactivity Control System 
RCSS Reactivity Control and Shutdown 
System 
RPS Reactor Protection System 
RPV Reactor Pressure Vessel 
Reactor Pressure Vessel 
RPVCS Conditioning 
System 
RSS Reserve Shutdown System 
SAR Safety Analysis Report 
SAS Small Absorber Sphere 
SBS Start-up Blower System 
sic Silicon Carbide 
TRISO Triple Coated Particle 
WMPS Water Make-up Purification System 
CHAPTER 1 
INTRODUCTION 
1 . 1 Introduction 
Nuclear energy has both benefits and risks . One of its many 
benefits is power genera t ion . Bec ause of the inherent risks , to 
generate power from nuclear energy, utilities must show that the 
reactor will operate safely without unduly exposing the public 
and the workers to health risks. One of the techniques used to 
ensure safety in nuclear installations is probabilistic risk 
assessment (PRA) . This is also variably known as probabilistic 
safety assessment (PSA). While regulators across the world are 
increasingly recognising PRA in l icensing nuclear reactors , in 
South Africa nuclear power plants have been licensed based on PRA 
for years . 
1.2 Background 
PRA was first used to assess the safety of nuclear reactors in 
1975 as part of the Reactor Safety Study commissioned by the 
United States Nuclear Regulatory Commission [l] . It has since 
gained use as a fundamental t ool for assessing safety of 
2 
installations worldwide. For example, PRA is used in such varied 
industries as the chemica l and aviation . 
The PBMR (PTY) LTD has just completed a Phase 2 PRA to assess the 
safety of the design [2] . Phase 2 PRA only assesses the safety 
impact of internal ev ents. External events have not been 
adequately analysed in the current design . It is standard 
practice to also determine the impact of external events on the 
safe operation of the plant. Example s of external events i nclude 
fire, earthquakes , flooding, aircraft crash, and tornadoes , etc. 
This study focuses on the fire assessment. The safety ass e ssment 
of other external events will need to be performed at some stage. 
1 . 3 Overv iew of PRA 
1u:~iv/ 
The main objectiv e of the PRA is to identify all possible way s 
that might result in plant personnel and the public being exposed 
to radioactiv e material and then determining their frequencies 
and consequences. Frequencies and consequences are then combined 
to determine the risk due to the various accidents. In 
pressurised water reactors (PWR) , it is sufficient to determine 
all scenarios, which might result in the reactor core me l ting. 
This is because substantial amounts of radioactiv ity are assumed 
to be released only after large scale destruction of the nuclear 
fuel. In the PBMR , however, core melt is highly unlikely because 
of the fact that PBMR is i nherently safe (this means that the 
occurrence of any ev ent would not result in exposure of the 
3 
public to the radioactive material ) , the negative temperature 
coefficient and the temperature tolerance of the fuel are such 
that sub-criticality without forced cooling can be achiev ed and 
maintained without active engineered systems or operator action 
for many hours. However , there are scenarios that may result in 
radioactivity reaching plant personnel and the public [2] In 
performing PRA, it is i mportant t o ensure that all possible ways 
in which this can happen are identified. The procedure for 
conducting a PRA can be found in Chapter 2 . 
The probabilistic risk assessment method was introduced to 
evaluate the interaction between potential system failure and the 
resulting health impact . It is an analytical technique 
integrating a wide spectrum of design and operational aspects of 
a specific faci l ity [9, 10 ] . 
1. 3 .1 Why is fire of concern 
Historically, fire has been found to contribute significantly to 
risk of nuclear power . Events [8] have shown that fire in nuclear 
power plants although rare, could increase the risk due to their 
operation . The special nuclear boundary conditions (like, the 
absolute confinement of radioactivity) impose severe requirements 
on the fire protection and the fire fighting measures. Since the 
Browns Ferry Fire major efforts have been made to further reduce 
the risk of fires in nuclear power plants [19] . 
4 
Fire also tends to induce multiple failures. Nuclear power plants 
must demonstrate t hat t he releases of rad ioactivity wil l still be 
within the limits , given a fire in any area of a plant . Although 
fire can occur anywhere within the plant, the risk will be 
greatest in certain crit i cal fire areas [5]. 
1. 3 .2 Historical Evidence 
There have been a few incidences of fire in nuclear power plants . 
Al though some of them were minor, others threatened the safe 
operation of the plants. A few examples are discussed here. 
The Browns Ferry Fire accident is one example of why fire is a 
major risk [8]. The accident occurred during maintenance when one 
of the maintenance personnel lit a candl e in the cabl e room. The 
cables caught fire. From these accidents it also became evident 
that critical fire areas in power plants apart from areas 
containing the safety equipments are areas containing electrical 
equipments such as the cables, motors, transformers , switches, 
batteries, generators, b l owers, electrical busbars and breakers 
and electrical cabinets. 
1.3. 3 Causes of fire 
Experience has shown that maintenance activities could lead to 
fire accidents as explained above by the Browns Ferry Fire. 
Electrical equipments can also cause fire. Electrical fires can 
be explained by Ohm's law, which is expressed as: 
5 
(V = IR) where, 
Vis the voltage drop in volts, 
I is the current in amperes, and 
R is the resistance in ohms 
The Power (P) (in Watts), in an electrical wire can be expressed 
as: 
p VI, but, since V IR 
Then P I 2R 
From this equation it follows that when current increases and 
resistance also increases by the equation: 
The power in the cables increases, and this then results in 
electrical cables becoming hot and there after they start 
burning. This is how fire occurs in electrical equipments. 
At this stage in the PBMR PRA, the impact of fire on the 
risk to the public due to the operation of the plant has 
not been investigated. It is therefore very important that 
a fire analysis be incorporated into the PBMR's risk model. 
6 
1.4 Background 
The importance of fire as a potential initiator of multiple-
system failures took on a new perspective after the cable tray 
fire at Browns Ferry plant in 1975, where a fire was ignited 
during maintenance as a result of one of the maintenance 
personnel lighting a candle in the cable room [8] 
Fire analysis involves quantifying the frequency of core melt or 
the ability of a plant to reach a safe reactor shut down, and 
containment failure given a fire incidence anywhere in a plant. 
The frequencies and consequences of fire in each area of a plant 
and the impact of these consequences on the safe shut down 
capability of the plant are assessed [5]. 
Fire analysis process requires knowledge of several important 
aspects of a fire incident like: ignition, progression, 
detection, and suppression, and characteristics of various 
materials under fire conditions as well as the plant safety 
functions and their behaviour under accident conditions [12). 
Fire research is a multidisciplinary effort, which covers a large 
spectrum of topics like: the physics of combustion, the behaviour 
of flames in compartments, and the characteristics of fire 
detector responses [8]. 
To perform an assessment of fire impact on risk knowledge of the 
plant is required . The description of the plant is provided in 
the following section. 
7 
1. 4 . 1 The Pebble Bed Modular Reactor Power plant 
The Pebble Bed Modular Reactor (PBMR) plant is different from 
ordinary Light Water Reactors (LWR) systems in that : 
• It is inherently safe, meaning that even a major accident 
such as a Depressu rized Loss of Forced Cooling (DLOFC) the 
reactor will still be able to sustain itself without any 
operator actions by natural convection where heat will be 
transferred from the fuel in the core to the reflectors and 
from the reflectors and all other systems to the reactor 
wall and then to the ultimate heat sink. 
• The negative temperature coefficient and the temperature 
tolerance of the fuel are such that sub-criticality without 
forced cooling can be achieved and maintained without 
active engineered systems or operator action for many 
hours. 
The PBMR is a graphite moderated helium cooled reactor, which 
uses a Brayton direct gas turbine cycle to convert the heat, 
which is generated in the core by nuclear fission . The heat 
generated in the core by the fission process is then transferred 
to the coolant gas. The gas is then transported to the turbine 
where it is converted into electrical energy . Figure 1-1 General 
layout of the Reactor Unit and the Power Conversion Unit (October 
2000). The use of Helium eliminates the possibility of a change 
of phase of the coolant (i . e. liquid to vapour) under accident 
conditions . Accident analysis is greatly simplified and 
uncertainties reduced by the absence of two-phase flow. 
Figure 1-1 General layout of the Reactor Unit and the Power 
Conversion Unit (October 2000). 
1-- 1 
The basic reactor power generation system consists of: 
• Reactor Unit; consisting of the Reactor Pressure Vessel, 
Core Structures , and Reactivity Control and Shutdown 
Systems. 
• Power Conversion Unit; consisting of High Pressure Turbo-
uni ts, Low Pressure Turbo-uni ts , Power Turbine , Generator 
System, Recuperator, Coolers, Hot Pipe and Valve System. 
9 
• Main Support Systems; consisting of Fuel Handling and 
Storage System, Helium Inventory Control System and Active 
Heat Removal System [PBMR SAR-Section 3, 2001]. 
There are a number of safety systems by which the reactor 
activities can be controlled. These include activities such as 
the heat removal, and the control of heat generation. For heat 
removal systems such as the Reactor Cavity Cooling system (RCCS), 
Start-up Blower System (SBS), and the Core Conditioning System 
(CCS) are used. For the control of heat generation systems such 
as the Reactivity Control System (RCS), and Reserved Shutdown 
system (RSS) are used. 
The function of the RCCS is to remove heat from the reactor 
during normal operation and shutdown. The system also removes 
decay heat during the loss of the heat transfer functions of the 
Power Conversion Unit (PCU) The objective is to prevent the 
reactor vessel, Reactor Pressure Vessel (RPV) supports, 
instrumentation and concrete walls from exceeding from exceeding 
their design temperature limits . 
The CCS maintains the RPV at desired temperatures (280 °C to 300 
°C) during normal power operation, it also removes core heat 
during maintenance operations when the SBS cannot be used and it 
can be used to recover the core from very hot conditions (up to 
1200 °C) in the event of a PCU trip followed by failure of the 
SBS blowers to operate . During maintenance it is used to maintain 
10 
the core at the required temperature although the SES blowers 
will be used to initially cool the core down to low temperature 
maintenance conditions. 
The RCS consists of a control rod system, and the RSS consists of 
the shutdown rods and the Small Absorber System (SAS) Both the 
RCS and the RSS are independently able to transfer the reactor to 
a sub-critical state, and keep the reactor subcri ti cal for an 
indefinite period at normal operating temperature. Both systems 
combined must allow the reactor to be in a sub-critical state 
below 50 °C indefinitely. 
uBRA'ftvJ 
Figure 1-2 The Pebble Bed Modular Reactor fuel elements 
FUEL ELEMENT DESIGN FOR PBMR 
5mm Graphite layer 
Coated particles imbedded 
in Graphite Matrix 
Dia. 60mm Pyrolytic Carbon 
Fuel Sphere Silicon Carbite Barrier Coating Inner Pyrolytic Carbon 
Porous Carbon Buffer 
Dia. 0,92mm 
Coated Particle 
Dia 0,5mm 
Uranium Dioxide 
Fuel 
The PBMR uses spherical fuel elements . Th e PBMR reactor consists 
of graphite spheres situated at the centre of the reacto r core 
s u rrounded by fue l spheres which also con tain a layer of graph ite 
and fuel particles known as Triso particles because of the three 
layers covering them (i.e. the inner pyrolytic carbon layer, the 
pyrolytic silicon carbid e layer, and the outer pyrolytic carbon 
layer). 
1.5 Purpose of study 
Fire PRA is important for ident ifying t h e possible consequences 
of fire , which could prevent t h e safe, shu t down of a nuclear 
power plant [8] . The purpose is also t o investigate the 
12 
methodology and further apply it to quantify the frequency of 
fire in the PBMR . 
The Regulatory guide [3 ] mandates fire prevention and protection 
programs . It also requires that fire contribution to the 
radiological consequences be included in the probabilistic 
assessment . At this stage in the PBMR PRA development, fire will 
need to be incorporated into the PBMR's probabilistic risk model . 
The importance of this study lies in using the probabilistic 
methods to calculate the frequency of reaching a safe reactor 
shutdown given that fire occurred anywhere in the PBMR power 
plant. 
1.6 Objectives 
The objectives of the study are: 
I. To present the methodology of the study. 
II . To use the methodology to quantify the frequency of 
the safe shutdown capability of the plant in case 
of a fire accident. 
III . To eva l uate the impact of fire on the safety of the 
PBMR power plant. 
1.7 Outline of chapters 
Literature survey on fire and general fire analysis 
methodologies, their ignition, how they spread, their 
13 
detection, suppression and mitigation are discussed in 
Chapter 2. 
Chapter 3 focuses on the step - by- step process of a fire 
analysis. The end result is the frequency of reaching a 
safe shutdown in case of fire. Uncertainty calculation is 
not in the scope of the study. 
Chapter 4 discusses the results of the study. The summary, 
conclusions and recommendations are discussed in Chapter 5. 
CHAPTER 2 
LITERATURE SURVEY 
2 . 1 Revi ew of Lite rature 
1. The book NFPA 805 [20), specifies the minimum fire protection 
requirements for existing light water nuclear power plants during 
all phases of plant operation, including shut-down, degraded 
conditions, and decommissioning . The book also prov ides 
information on methods of maximising defence in depth during a 
plant fire, goals and objectiv es of nuclear safety . 
The fire protection standard is based on the concept of defence 
in depth. Protecting the safety of the public, the env ironment, 
and plant personnel from a plant fire and its potential effect on 
safe reactor operations is paramount to this standard. Defence in 
depth is achieved when an adequate balance of each of the 
following elements is provided; ( i) preventing fires from 
starting, (ii) rapidly detecting fires and controll i ng and 
extinguishing promptly those fires that do occur, thereby 
limiting fire damage, (iii ) providing adequate level of fire 
protection for structures, s y stems, and components impor tant to 
safety so that a fire that i s not promptly extinguished wi ll not 
prev ent essential plant safety funct i ons from being performed . 
15 
The safety goals include (i) the provision of reasonable 
assurance that a fire during any operational mode and plant 
configuration will not prevent the plant from achieving and 
maintaining the fuel in a safe and stable condition; (ii ) to 
provide reasonable assurance that a fire will not result in 
radiological release that adversely affects the public, plant 
personnel, or the environment; (iii) to provide reasonable 
assurance that loss of life in the event of fire will be 
prevented for facility occupants; (iv) to provide reasonable 
assurance that the risks of fire are acceptable with regard to 
potential economic consequences. 
The nuclear safety objectives during a fire event are: (i) 
capability of rapidly achieving and maintaining subcritical 
conditions, removing decay heat and inventory control functions, 
preventing fuel clad damage so that the primary containment 
boundary is not challenged; (ii) maintenance of the containment 
integrity and the limitation of the source term; (iii) to protect 
occupants not intimate with the initial fire development from 
loss of life and improve the survivability of those who are 
intimate with the fire development as well as to provide adequate 
protection for essential and emergency personnel; (iv) to limit 
the potential property damage due to fire to acceptable levels, 
also the potential business damage due to fire is also limited to 
acceptable levels as determined by the operator. 
16 
The book also outlines a performance criteria by which fire 
protection features shall be capable of providing reasonable 
assurance that, in the event of fire the plant is not placed in 
an unrecoverable condition. This shall be demonstrated by: (i) 
insertion of the negative reactivity by the reactivity control to 
achieve and maintain sub-critical conditions such that fuel 
design limits are not exceeded; (ii) capability of inventory and 
pressure control to control coolant levels such that subcooling 
is maintained for PWR and the capability of maintaining or 
rapidly restoring reactor water level above top of active fuel 
for a BWR such that fuel clad damage as a result of a f i re is 
prevented; (iii)removing sufficient heat from the reactor core or 
spent fuel such that fuel is maintained in a safe and stable 
condition; (iv) capability of vital auxiliaries to provide 
necessary auxiliary support equipment and systems to assure that 
the systems required under (i ) , (ii ) , (iii), and (v) are capable 
of performing their required nuclear safety functions; (v) 
process monitoring for providing necessary indication to assure 
the criteria addressed in (i ) through to (iv) hav e been achieved 
and are being maintained. 
2. Lees F.P, in his book [12) defines fire in the following way: 
"Fire or combustion, is a chemical reaction in which a substance 
combines with oxygen and heat is released. Fire occurs when a 
source of heat comes in contact with a combustible material. 
17 
The development of fire depends on the situation of the fuel . 
That is, the stacking of materials in closely spaced vertical 
piles tends to encourage fire spread. 
Fire normally grows and spreads by direct burning, which results 
from impingement of the flame on combustible materials, by heat 
transfer or by travel of burning material. 
The three main modes of heat transfer from fires are: conv ection, 
conduction, and radiation. Conduction is particularly important 
in allowing heat to pass through a solid barrier and ignite 
material on the other side. Radiated heat is transferred directly 
to nearby objects and does not go preferentially upwards and 
crosses open spaces. Convection heat transfer involves the 
transfer of heat from one material to the next different 
material. Most of the heat transfer from fires is by convection 
and radiation . 
Electrical equipment is widely used in power plants and may be 
the source of ignition unless close control is exercised. The 
basis for the control of electrical equipment to prevent its 
acting as a source of ignition is the classification of the plant 
according to the degree of hazard." 
This information is important in that it provides knowledge about 
how fires occur, the propagation of fire from one place to 
another and also some adv ice on how to place materials or 
equipments in power plants (not placing materials in stacks or 
18 
piles aids in minimizing fire growth). It further provides 
information on areas that could be classed as fire areas or areas 
in which fires are expected in power plants. 
3. The paper (Schneider U.) [24] Outlines how difficult it is 
to prevent fires in nuclear power plants because of the 
combustible materials, which must be used for principle reasons, 
that is, power plants cannot avoid the application of such 
materials or media in certain areas or components of the plants. 
According to the paper , a detailed study on the fire occurrence 
rate in Pressurized Water Reactors (PWR) indicated that the 
occurrence rates of American and German reactor facilities were 
of comparable size. Also the paper states that it is indicated 
that the 95% fractile of the occurrence rate per year and 
facility is of the order of 3 to 4x10 -1 , the appropriate mean 
value is of the order of 1 . 1x10-1 • This means that one can assume 
that in average a fire in a nuclear power plant occurs every ten 
years. With respect to the type of materials involved these 
observations were made : roughly 1/3 of all fires occurred with 
oil or hydrogen, 1/3 was related to other materials. These data 
indicate the great importance of cable fires and the strong need 
for fire prevention . 
4. Fischer et al . [7] present the following v iews regarding fire: 
" In the design process the fire protection engineer needs to 
frequentl y work closely with the plant design team from the 
19 
beginning. The conceptual influence of fire protection 
considerations starts with the plant layout; the separation as 
far as possible of unavoidable fire loads and safety s y stems; 
good accessibility; separation between safety s y stems; and the 
installation of effective barriers. 
The need for the reduction of fire loads is generally 
acknowledged, but in many instances it is a question of material 
properties in which the designer has only limited choice. The 
removal of unnecessary combustible materials is generally 
practiced . " 
5 . The paper ( Gregory et al) [9] describes how important a 
combustion analysis of an entire facility is, in determini ng the 
transient pressure and heating loads on the containment or 
ventilation structures and safety related equipment, to plan 
emergency evacuation, to validate existing fire prot ection 
programs or suggest changes, and to integrate fire safety with 
other safety analysis requirements . 
It then concludes that fires in reactor systems may be initiated 
and then propagate during transient releases of heat, carbon 
monoxide and hydrogen and therefore nuclear reactor crit i cality 
transients may l ead to severe accidents . 
6 . The paper (Himanen et al ) [8] describes what the fire analysis 
method required for the TVO units I and II, which is common in 
all other methods. It required that all 1400 rooms of the plant 
20 
be visited, and it was conducted in connection with the annual 
fire inspection made during and immediately after its refuelling 
outage in 1988. The paper further outlines the probability of 
ignition in a room depends on the presence of persons and on the 
amount of electrical and mechanical equipment in the room, and 
that the propagation of fire is a function of numerous factors 
for example, the amount and location of combustible material, 
ventilation and many others. The total fire frequency of the 
plant was estimated on the basis of the experience gained in the 
two units of the TVO plant and similar plants in Sweden. 
7. According to this book (Safety Series) (22] a Safety Analysis 
Report of any power plant should provide a description and a 
safety analysis of the fire protection system in that particular 
plant including information on procedures and maintenance 
activities. The section on the Design for Internal Fire 
Protection should describe the design requirements for fire 
protection inside the facility. According to this book this 
should also include passive features, such as isolation, 
separation, selection of materials, the building layout and 
zoning, the location of fire barriers and the safety system 
layout and protection including the separation of safety related 
redundant systems. NWU 
1.. . IBRARY 
This is a book which is accepted and used world wide by plants 
with hazardous materials and it is recommended by the nuclear 
21 
regulators internationally that organizations/plants dealing with 
hazardous mater i als perform up to t he standards set by this book . 
8 . Licensee Event Reports (LER), Computer Aided Regulatory 
Library (CARL) 
Power plants are supposed to report all types of events big or 
small , occurring during any mode of operation. The reason f or 
this is so that power plants should learn from each other's 
experiences and to avoi d any event s that may lead to accidents . 
The fol l owing are exampl es of t h e Licensee Event Reports (LER) 
from the Computer Aided Regulatory Library (CARL) . In these 
reports all types of events t hat occu rred in dif f erent power 
plants were reported with the date of occu rrence and the causes 
of these event s and also the p oss ible solutions to these ev en ts 
for future re f erences. 
I. ANO 1 - Event Date: 10/17/1996 - LER Number: 1996 - 009 -
Cracked Weld in an Oil Line on a Reactor Coolant Pump: 
A fire was di s covered in insulation around the Main 
Feed wa t er nozzle ring on "B" Once Through St eam 
Generator (OTSG) during heat up of the Reactor Coolant 
Sy stem (RCS ) . A weld located in the discharge line of 
a Reactor Cool ant Pump (RCP) mot or oil l i ft pump had 
cracked due to a fabrication defect . The fail u re, 
believed to have occurred at the start of the outage , 
resulted in oil being introduced onto the insulat i on . 
Oil on the insulation allowed a wicking effect that 
22 
reduced the auto-ignition point of the oil to a value 
lower t han the documen ted va l ue . The fire originat ed 
when the RCS temperature was approximately 439 degrees . 
Application of a light water fog from a fire hose 
extinguished the fire app roximately 16 minutes after it 
was discovered. A Notification of Unusual Event was 
declared when the fire was n o t extinguished within 10 
minutes . The plant returned to cold shutdown 
conditions to evaluate damage . Other than some minor 
damage to insulation , the fire did not damage any 
systems or comp onents. Enhancements were made to the 
oil col lection systems of all RCP motors , and damaged 
insulation wa s repaired or replaced prior to t he 
subsequent heat up. 
II. BEAVER VALLEY 2 - Event Date: 04/02/1998 - LER Number: 
1998-005 Inadequate Fire Protection Safe Shutdown 
Analysis for Boric Acid to Boric Acid Blender Valve, 
2CHS*FCV113A: On 4/2/98, a fire protection safe 
s hutdown review for a proposed modification to the 
Boric Acid to Boric Acid Blender Valve, 2CHS*FCV113A, 
revealed that the fire protection safe shutdown 
analysis for t h e valve did not i dentify all cables and 
supporting equipment for operation of the valv e. As 
discovered , errors exis t in the va l ve control circu it 
analys i s and the description in the Beaver Valley Power 
Station Unit 2 (BVPS-2) Updated Final Safety Analysis 
23 
Report (UFSAR) of equipment operability following a 
fire in fire areas CPl, CV3, PA3, PA4, PA5, SBl, SB3, 
SB4, and TBl. Due to these errors, required actions for 
boration via the boric acid tanks specified in the 
analysis and the plant operating manual are not 
adequate for fire areas CV3 and PA5. As such, following 
a fire event, the valve and the required boration flow 
path may not be available. This condition does not 
comply with the fire protection evaluation, as 
described in Section 9.5A of the BVPS-2 UFSAR. As 
understood, this condition does not apply to BVPS-1 . 
This event resulted from oversight of the missing 
support cables and equipment for the valve while 
developing the list of electrical cables in specific 
designated fire areas for the fire protection safe 
shutdown analysis. Prior to entry into Mode 4 from the 
current Unit 2 shutdown, the control circuit of 
2CHS*FCV113A will be modified so that the valve will 
fail to the safe (open) position, in the event of fire 
damage where the redundant boration flow path is 
affected. By 6/30/98, the valve control circuit 
analysis will be documented. An extent of condition 
review for Unit 2 of this occurrence will be completed 
prior to Unit 2 entry into Mode 4. A revised Unit 2 
Fire Protection Safe Shutdown Report will be issued, by 
10/31/98. By 10/31/98, the condition report of this 
24 
event will be made part of the required readi ng and 
d i s tribution to Nuclea r Engineering Design pers onnel . 
III . BEAVER VALLEY 1- Inadequate Pre - Fire Plan Procedure 
Guidance Resulting in Potential Damage to Emer gency 
Diesel Generators : On April 9, 1999, with Unit 1 in 
Mode 1 at approximately 98% power, an ongoing 
Engineering Fi re Protection Safe Shutdown self-
assessment identified that operator actions in certain 
Unit 1 pre - fire plan and alternate safe shutdown 
procedures were inappropriate as they would place t he 
Unit outside the design basis of the Fire Protection 
Program for Appendix R Safe Shutdown capability . These 
actions would de-energize power to the Emergency Diesel 
Generator (EDG) Building Area Ventilation Sy stem 
Discharge Dampers lVS-D - 22 - lA and lB, and could have 
prevented proper v entilation for the EDGs and the 
fulfilment of their safety function to satisfy the 
Appendix R Safe Shutdown requirements . Non-emergency 
one-hour notification of this discovery was made to the 
NRC, at approxi mately 1846 hou rs on April 9, 1999, 
pursuant to 10 CFR 5 0 . 7 2 ( b) ( 1 ) These procedural 
inadequacies are not applicable to the corresponding 
Unit 2 procedures . The apparent cause of this event is 
personnel error attributed to failure to identify that 
the manual actions specified in the procedures would 
result in closure of the EDG v entilation exhaust 
25 
dampers . A contributor to this personnel error was 
inadequate knowledge of the design of the subject 
ventilation dampers, which are spring return to close 
on loss of power . Following event determination, 
operating shift briefings were conducted to ensure 
awareness of the subject procedural inadequacies 
pending revision of the procedures. Unit 1 subsequently 
entered Mode 5, at approximately 2235 hours on April 
13, 1999. Appropriate rev isions to the affected 
procedures hav e been implemented to ensure that power 
to lVS-D-22-lA and lB is not de-energized by the 
procedures. 
2.2 General fire analysis methodology in power plants 
Fire analysis in nuclear power plants presently involves 
following a number of procedures. It also requi res knowledge 
about fires, material properties , combustion, plant design and 
layouts, cable locations, and the location of safety equipments 
[ 9] . 
The purpose of the analytical method developed is to identify a 
list of the dominant accident sequences that are initiated by 
fire and then to assess the frequency of occurrence for each . 
This process requires information about several important aspects 
of a fire (ignition, progression , detection, and suppression, 
characteristics of materials under fire conditions) as well as 
26 
the plant safety functions and their behaviour under accident 
conditions. Considerable uncertainties exist in the analysis 
because of gaps in the required knowledge [13). 
The process of fire analysis can be divided into three main 
steps. The three steps are: 
I. A hazard analysis, 
II. A fire propagation analysis, which is somewhat 
analogous to a component-fragility analy sis, and 
III . A plant and system analysis 
The hazard analysis develops the frequency and magnitude of the 
externally imposed stress, where "stress" is in terms of 
potential fire induced accident sequences. 
The propagation analysis investigates the resistance of the plant 
to fire damage by studying the propagation of the fire and the 
effectiveness and timing of suppression. 
The last analysis evaluates the response of plant systems to the 
accident sequence triggered by the fire; it considers core damage 
[13) . 
CHAPTER 3 
METHODOLOGY 
The methodology to be followed in fire risk assessment invo l ves 
three steps. These steps were briefly introduced in the previous 
section and are discussed in detail in this section. The steps 
are: 
I. Fire hazard analysis, 
II. Fire propagation analysis, and 
III. Plant and s y stem analysis 
3.1 Fire Hazard analysis 
The purpose of the fire-hazard analysis is to identify the 
critical plant locations that are important to the fire-risk 
analysis. This is analysis is carried out in two steps. The 
first step is a location screening exercise, which identifies the 
critical plant areas. The second step involves the estimation of 
the fire frequency. 
3.1.1 Location Screening 
Theoretically, one must determine all areas with the potential 
for fire starting. However, this is practically impossible and 
28 
is not necessary as some areas may not be important. This step 
of the procedure allows for unimportant areas to be screened out 
while retaining areas that contribute significantly to risk. The 
locations containing the fire vulnerable safety equipment are 
identified. The loss of all equipment at that location is then 
postulated . If it is found that an initiating event (Loss of 
Coolant Accident (LOCA) or transient) will not occur, the 
location is eliminated from consideration . Given a LOCA or 
transient, a number of safety functions are required for safe 
shutdown. If the loss of all equipment in the location of 
interest prohibits the performance of any or all required 
functions, the location is tabbed for further analysis. 
This screening method, described by Kazarians and Apostolakis 
(1981 ) was employed in the fire risk portions of the Zion 
(Commonwealth Edison Company , 1981) and Indian Point studies 
(PASNY, 1982 ) In these analy ses, the fire induced loss of 
control of safety s y stems is judged to dominate fire induced 
hardware losses, and therefore the critical locations that are 
investigated contain electrical cables and/or switchgear fo r many 
safety and non safety systems. 
Given this assumption, and a fire that has caused an initiating 
event, it is then important to determine whether the same fire 
can induce failures that will prevent: 
1. Reaching and maintaining a condition of negative 
reactiv ity . 
29 
2. Removing decay heat. 
3. Moni taring and controlling the inventory and pressure of 
the reactor-coolant system (RCS). 
Because of the fail-safe logic of the reactor scram circuitry , it 
is then assumed that the reactor trip is successful. Therefore 
critical fires wi ll affect the implementation of actions 2 and 3. 
If no one fire can do this, the fires that can prevent either 
action 2 or action 3 are considered. However these fires are less 
likely to be significant contributors to risk, because at least 
one independent failure in an unaffected safety system is 
required for (core damage and) radionuclide release to occur. 
3 .1.2 Fire Occurrence Frequency 
Once the locations where a fire has the potential to initiate an 
accident sequence leading to a release of radionuclides have been 
identified, it should be determined how often these fires occur. 
The rate of occurrence can be established from the historical 
record. Only the plant operating histories are suitable for 
assessing the risk from plants at power. 
Apostolakis and Kazarians (1980) model the frequency of fires for 
various compartments, using a probability of frequency framework 
to consistently treat the uncertainties. Starting with broad 
prior distributions to model their weak state of knowledge they 
employ the statistical evidence of fires in a specific type of 
30 
reactor (e . g . light water reactor) and use Bayes' Theorem, to 
derive the fire frequenc i es in specific rooms of the building . 
The procedure can also be used to update the given distributions 
for fire frequencies, when further reactor experience is 
accumulated. 
3 . 1. 3 Fire Propagation Analysis 
Event trees are used to d ivide the fire model into four elements: 
Ignition, Detection, Suppression , and Propagation. Each element 
wi l l head a column of the event tree. The fire is assumed to 
start in one component and potentially propagate to the next . 
. NWU · 
3.1.4 Plant System Analysis J.LIBRARYJ 
Once the frequencies of fire-induced c omponent losses are 
asses s ed, it is possible to estimate the frequency of fire 
initiated accident sequences leading to c ore damage (which is not 
the case for t h e PBMR since core dama ge is almost impossib le 
because of the fact t hat the PBMR has a negative temperatu re 
effect ensuring that al l core structures never reach temperatures 
at which they start losing their integrity including graphite) . 
Like other initiating events , separate event trees are 
cons t ructed for fires bec ause the operator , rather than automat ic 
actions, may be responsible for shutting down the plant in 
response to a fire . This is often done by simply modifying the 
front end of existing event trees for other initiating events to 
spec ialize them for fire s . The postulated f ire can be assumed to 
31 
coincide with another initiating event, in which case the 
original event tree structure would be re t ained. (NB. Howev er , if 
fires are to be treated as a separate event, care should be taken 
that data from which basic component-failure rates are determined 
do not double-count these failures from fires). 
Disadvantages of the Method 
This method however detailed and like many others has 
disadvantages. One of the disadvantages is the potential for 
omitting fire sequences leading partway to core meltdown but 
requiring additional component failures to result in the top 
event. Another disadvantage is that this method does not account 
for the possibility that the fire-caused simultaneous failures of 
many instruments and/or nonsafety s y stems may initiate accident 
sequences . 
Conunents 
This method was chosen because compared to the other methods it 
is more detailed and more explanatory than other methods though 
it requires a lot of details about the plant, but it contains 
more familiar information than the other methods . It may not be 
possible to do all that this method requires, the reason being 
that we do not have a built plant yet and ev ery thing about the 
plant keeps changing so there is missing information and some of 
the things haven't been done yet . 
32 
3.2 Application of the methodology 
3.2.1 Fire Hazard Analysis 
This methodology has been mostly for Light Water Reactors (LWR), 
Pressurized Water Reactors (PWR), and Boiling Water Reactors 
(BWR) . The methodology does not include some of the steps done 
here, but the methodology had to be altered because as mentioned 
previously the PBMR plant is still in the design phase . 
3.2.1 .1 Identification of Safety Functions and their Equipments 
Safety functions were identified in order to make a list of 
related systems so as to know which systems will be more 
vulnerable and where to locate them in the building layout. 
Safety functions are processes by which power plants control the 
reactor activities to ensure that there are no accidental 
releases of radioactiv ity to the env ironment. 
These are activities such as: heat removal/ cooling, control of 
heat generation (that is the control of the fission chain 
reaction) . 
There are systems involved in fulfilling the safety functions and 
these systems also have sub-systems that act as their support 
systems. 
Table 1 Safety Systems and their Support Systems contains the 
safety functions that are important for reaching a safe reactor 
33 
shutdown in case of a fire accident; it also contains the related 
safety systems and their support s y stems. 
Table 1 Safety Systems and their Support Systems 
Safety Function Safety System Support System 
Heat removal SBS Electricals, Water 
circuits, Back-up 
ccs cooling tower (for 
RCCS & CCS) I 
RCCS Compressed air (for 
CCS). 
Reactivity Control RCSS: RCS Electricals 
RSS 
Fire tends to induce/introduce inter-sy stem dependencies in power 
plants. 
34 
Table 2 Safety System Components and their Support Systems 
(System Dependencies) be l ow show the safety system components and 
their inter-system dependencies . 
35 
Table 2 Safety System Components and their Support Systems 
(System Dependencies) 
Support Systems 
Safety System 
components 
1.Core Fuel-RCS 
18 Control Rod Power 
Systems which plant 
consist of the RPS EPS electrical 
following (per system 
system): 
Control rod .,, .,, 
Control Rod .,, .,, .,, 
Drive Mechanism 
(CROM) with 
position 
indicator 
Rod Emergency 
Shock Absorber 
2.Core Fuel-RSS 
17 Mutually 
Independent SAS 
Systems which 
consist of the 
following (per 
system): 
Gas blower .,, 
(shared) 
Storage 
container 
Discharge 
container 
Storage .,, .,, .,, 
container valve 
36 
3.Core Fuel-CCS ACS- ACS- Power 
Auxiliary Back-up 
plant 
closed cooling 
WMPS EPS ocs electri 
circuit tower 
cal 
closed 
system 
circuit 
Gas mixing 
chamber 
Gas-to-gas heat 
exchanger 
(Recuperator ) 
Gas-to-water 
heat exchanger 
(stainless 
steel plates) 
Buffer circuit .; .; 
heat exchanger 
(stainless 
steel plates) 
Buffer circuit .; 
expansion tank 
Two buffer .; .; .; 
circuit pumps 
Buffer circuit 
pump outlet 
NRVs 
3 X 50% blowers .; .; .; 
3 X blower .; .; .; 
inlet motor 
operated valves 
3 X blower .; .; .; 
return line 
motor operated 
valves 
Recuperator by- .; .; .; 
pass motor 
operated valv e 
4.SBS EPS ocs Power 
plant 
37 
electrical 
system 
2 100% .,, .,, .,, X 
blowers 
2 X blower .,, .,, .,, 
outlet motor 
operated valves 
5.Core Fuel- ACS-Open ACS- WMPS EPS ocs Power 
RCCS, Train A circuit Back-up plant 
cooling electrical 
tower system 
closed 
circuit 
Heat exchanger .,, 
(titanium 
plates) 
Expansion tank .,, 
Expansion tank .,, .,, .,, 
WMPS motor 
operated valve 
Pump .,, .,, .,, 
Pump outlet NRV 
Standpipes .,, .,, .,, 
Anti-siphoning 
devices 
Rupture discs 
Temperature , .,, .,, .,, 
pressure and 
level 
indicators 
Back-up cooling .,, .,, .,, 
tower supply 
motor operated 
valve 
Back-up cooling .,, .,, .,, 
tower return 
motor operated 
valve 
38 
Standpipe v v v 
isolation motor 
operated valves 
6.Core Fuel-
RCCS, Train B 
Heat exchanger v 
(titanium 
plates) 
Expansion tank v 
Expansion tank v v v 
WMPS motor 
operated valves 
Pump v v v 
Pump outlet NRV 
Standpipes v v v 
Anti-siphoning 
devices 
Rupture discs 
Temperature, v v v 
pressure and 
level 
indicators 
Back-up cooling v v v 
tower supply 
motor operated 
valve 
Back-up cooling v v v 
tower return 
motor operated 
valve 
Standpipe v v v 
isolation motor 
operated valv es 
7.Core Fuel -
RCCS, Train C 
39 
Heat exchanger .,, 
(titanium 
plates) 
Expansion tank .,, 
Expansion tank .,, .,, .,, 
WMPS motor 
operated valves 
Pump .,, .,, .,, 
Pump outlet NRV 
Standpipes .,, .,, .,, 
Anti-siphoning 
devices 
Rupture discs 
Temperature, .,, .,, .,, 
pressure and 
level 
indicators 
Back-up cooling .,, .,, .,, 
tower supply 
motor operated 
valve 
Back-up cooling .,, .,, .,, 
tower return 
motor operated 
valve 
Standpipe .,, .,, .,, 
isolation motor 
operated valves 
40 
3.2.2 Location Screening 
The following locations were screened according to their 
containment of the fire vulnerable safety equipment. All other 
areas that were not included in the table below were screened out 
since they did not contain any fire vulnerable safety equipments. 
The table below is a description of the building layout . The 
first number represents the level of the floor, the negative sign 
means that the particular level is below the ground level, zero 
is the ground level, and the levels with no signs mean that those 
levels are above the ground level, the last numbers represent the 
room numbers. This is a list showing all the safety equipment and 
it also shows what these equipments are housed with in all the 
different rooms of the plant. The cable locations have been 
included, electrical switchgear rooms, generator room, control 
rooms, battery rooms, and the power electronics room because 
according to literature, these areas are areas, which have proved 
to be likely fire areas. ,~wu j 
LIBRARY 
41 
Table 3 Building Layout 
Area no Area List of equipment Maintenance 
Description in the area tasks in the 
(Level no area 
and room 
no) 
-18 : 32 Cable Duct 
Cable Duct Cables None 
-18:33 Cable Duct 
-18:34 Cable Duct 
-18 :35 Cable Duct 
-18:29 PCU Cavity PCU floor area: 
Pre & Intercooler . 
Waste Water Sump. Check sump 
pump s ystem 
functioning 
Intercooler Intercooler Inspection of 
outer casing 
for cracks , 
leaks 
Pre-cooler Pre-cooler Inspection of 
outer casing 
for cracks, 
leaks 
Recuperator Recuperator Inspection of 
outer casing 
for cracks , 
leaks 
Recuperator SBS 
SBS va l ves 
Recuperator Bypass 
valves 
PCU Area HPT 
LPT 
42 
Power Turbine 
-14:45 Core Core unloading Replace drive 
unloading device (x2) 
Device 
Replace 
gearbox 
Replace shaft 
penetration 
Test flange 
seals 
Replace 
spindle with 
bearings 
Empty scrap 
collection 
casks 
Repair bad 
electrical 
connection 
FHSS valves Isolation valves Leak test on 
(x40) flange seals 
Replace valve 
drive 
Replace shaft 
penetration 
Replace 
inserts 
Service shaft 
penetration 
Counters (x2) Leak test on 
flange seals 
Repair faulty 
electrical 
connection 
Replace 
inserts 
43 
Sphere r i ser pipes None 
Valv e blocks None 
Gas ev acuation Rep l ace 
s y stems He shut se r vice k i t 
off v alv es (x44) 
Fork Lifts None 
Valv e maintenance None 
tool 
Lay down pallets None 
Core unloading None 
dev ice too l 
Scrap cask None 
empty ing dev ice 
RSS v alv es 
-14 : 51 RSS blowers RSS blowers 
- 9:46 Bottom of the Bottom of RPV NDT the bot tom 
RPV of the RPV for 
cracks, 
deterioration 
of material, 
creep 
RPV RPV 
Hot & cold 
manifold pipes 
RCCS 
-9:47 Working floor 
area ccs 
-4:53 ccs ccs 
RPVCS 
-4 : 54 Electrical 1 1 kv switchgear Inspect 
s witchgear Equipment 
400 kv s witchgear Replace 
components 
44 
Dry transformers 
Circuit breaker 
trolleys 
Replace components 
-4:104 Electrical 11 kv switchgear Inspect 
switchgear Equipment 
400 kv switchgear Replace 
components 
Dry transformers 
Circuit breaker 
trolleys 
Replace components 
0:63 Generator Generator braking Test brake 
switchgear equipment system 
Lighting board 
Excitation control Calibrate and 
panel check 
Generator 
main/back-up 
protection panel 
Generator 
measurement panel 
0:60 Control room Operator desks Maintain 
panels and 
VDUs 
Video monitors 
PEI displays 
RPS displays 
Printers Check printers 
0 : 59 Control PLCs & control Replace cards 
equipment equipment 
-----
45 
room 
Engineering Plant 
control station optimization 
Configuration 
updates on 
plant 
4:75 Electrical Generator busbars Test 
busbars & transformer 
breakers 
Generator circuit Check circuit 
breakers breakers 
Brake air-cooled Check contacts 
transformer 
Brake circuit 
breaker 
Earthing 
transformer & 
resistor 
Voltage 
transformer 
Excitation 
transformer 
Earthing switch 
Generator earthing 
equipment 
4:111 Battery room 110 VDC electrical Inspect 
1 protection batteries 
batteries (2 sets) 
Top-up 
electrolyte in 
batteries 
I 
NWU· Replace 
\LIBRARYJ individual 
battery cells 
Replace all 
battery cells 
(end of life ) 
46 
Nicad batteries Inspect 
batteries 
Top-up 
electroly te in 
batteries 
Replace 
individual 
battery ce l ls 
Replace all 
battery cells 
(end of life ) 
Ventilation fans 
4 : 110 Battery room 110 VDC electrical Inspect 
2 protection batteries 
batteries (2 sets) 
Top-up 
electroly te in 
batteries 
Replace 
individual 
battery cells 
Replace all 
battery cells 
(end of life ) 
Ni cad batteries Inspect 
batteries 
Top-up 
electrolyte in 
batteries 
Replace 
individual 
battery cells 
Replace all 
battery cells 
(end of life) 
Ventilation fans 
47 
4:71 Power DC chargers Maintain t his 
electronics equipment 
room 
UPS 
EMB cabinets, 
including test 
equipment 
PLC equipment 
RCS & PEI 
4 : 73 RCS room 
11:80 RPV RPV No access or 
maintenance in 
this area 
RCCS 
RSS 
RCS 
21 : 87 Working floor RCSS maintenance 
trolleys 
RCCS water level Maintain 
sensors switches/senso 
rs 
Controlled Block filters Replace 
ventilation filters 
exhaust air 
plant Centrifugal fans Replace fans 
Dampers and motors Inspect and 
calibrate 
dampers 
Maintain and 1 y early 
replace motors 
Controls Inspect and 
maintain 
controller s 
and actuator s 
General All rooms A/C duct i ng 
48 
Cooling water 
supply lines 
Electrical cabling 
to individual 
equipment 
3.2.3 Fire Occurrence Frequency 
To compute the fire occurrence frequency for different specified 
compartments of the plant, statistical data and historical 
experiences are used, that is the number of fires that occurred 
previously in each room of interest in all plants of a similar 
type plus their number of years of operation are combined and 
from that, the fire frequencies are computed. 
The Pebble Bed Modular Reactor (PBMR) is the first reactor of its 
kind and it has not been built ye t, therefore there is not enough 
data and no experience to do the required calculations, like 
computing the fire frequency. There are however other HTGRs and 
HTRs that have been operating for a number of years, but there 
hasn't been any historical fire experience data that has been 
reported yet from them. In this regard statistical data and 
historical experiences available from reactors such as the LWRs, 
PWRs and BWRs (NUREG-1150 Volume 1 , NUREG/CR-2300 Volume 2, 
NUREG/CR-4840 SAND 88-3102) were adopted for the PBMR. 
49 
This data is suitable in a sense that, all cables, control rooms, 
batteries, generators, and electrical switchgears, in all power 
plants irrespective of the type of plant are the same . 
50 
Table 4 Statistical Evidence of Fires in Light Water Reactors and 
the related fire frequencies. 
Area Number of fires Number of Fire Fr equency 
compartment years (per room 
year ) 
Control Room 3 68 1. 0 4 . lE-3 
Cable s preading 7 . 2E-3 
2 74 7 .3 
r o om 
Diesel generato r l.9E-2 
37 1600 . 0 
room 
Reactor b u ild i ng 22 84 7. 5 5 . 0E-2 
Auxiliary 3 . 4E-2 
43 6 73.2 
build ing 
El ectric al 
4 13 46 . 4 l . 3E-2 
s wi tch g e a r room 
Battery room 4 1346. 4 2 . 9E-3 
51 
The fire frequencies for the different rooms of the PBMR plant 
were stipulated in the following manner: 
Generic subdivisions like the diesel generator room, cable rooms, 
control rooms, electrical switchgear rooms, and the battery rooms 
were considered to be common to all types of plants and so the 
fire event data were used directly with the number of years of 
operating experience per subdivision to determine the frequency. 
Plant specific subdivisions like the reactor building, and the 
auxiliary building were considered on a room-by-room basis. For 
these rooms the frequency was obtained by dividing the frequency 
from data amongst all systems contained in the specific 
subdivision. 
The fire frequencies for the reactor building, and the auxiliary 
building were divided for each room in a particular building 
according to the number of systems (especially systems that can 
burn) in each room re spec ti ve l y. And so the sum of all system 
frequencies in a room constituted the frequency of the part icular 
room. 
52 
Table 3-5 Fire frequencies for different rooms in the PBMR plant 
Area no Description Area List of Fire 
of location Description equipment in frequency 
(Level the area (per room 
no & year) 
room no) 
-18:32 Cable rooms Cable Duct 
2.88E-3 
Cable Duct Cables 
-18 : 33 Cable Duct l.44E-3 
-18:34 Ca b le Duct 1.44E-3 
-18 : 35 Ca b le Duc t 1 . 44E-3 
0:63 Generator Generator Generator 
r o om s witchgear braking 
equipment 
Lighting board 
Exci t a tion 
control panel 1 . 9E-2 
Gen erator 
main/back-up 
protect ion 
panel 
Gen erator 
meas urement 
pane l 
0:60 Cont ro l Control Operato r de s ks 2 . 0SE-3 
r o oms room 
Video moni t ors 
PEI d i splays 
RPS d i s plays 
53 
Printers 
0:59 Control PLCs & control 
equipment e quipment 
room 
2 . 05 E- 3 
Engineering 
control station 
4:111 Battery Battery 110 VDC 
rooms room 1 electrical 
protection 
batteries (2 
sets ) 
1. 4 5E-3 
Nicad batteries 
Ventilation 
fans 
4:110 Battery 110 VDC 
room 2 electrical 
protection 
batteries (2 
sets ) l .45E-3 
Nicad batteries 
Ventilat i on 
fans 
-4:54 El ectrical Electrical 11 kv 
s witchgear switchgear switchgear 
rooms 
400 kv 
switchgear 
Dry 
6 . 5E- 3 
transformers 
Circuit breaker 
trolleys 
Replace 
components 
-4:104 Electrical 11 kv 6 .5E-3 
switchgear switchgear 
54 
400 kv 
switchgear 
Dry 
transformers 
Circuit breaker 
trolleys 
Replace 
components 
-18:29 Auxiliary PCU Cavity PCU floor area: 
building Pre & 
Intercooler. 
Waste Water 
Sump . 
Intercooler Intercooler 
Pre-cooler Pre-cooler 
Recuperator Recuperator 
Recuperator SBS 3.4E-2 
SBS valves 
Recuperator 
Bypass valves 
PCU Area HPT 
LPT 
Power Turbine 
-14 : 45 Reactor 5.55E-3 
building 
FHSS valves Isolation 
valves (x40) 
55 
Gas evacuation 
s y stems He shut 
off valves 
(x44) 
RSS valves 
-14:51 Reactor RSS blowers RSS blowers 
1.85E-3 
building 
-9 : 46 Reactor Bottom of 
building the RPV 
RPV RPV 
3.7 0E-3 
RCCS 
-9:47 Reactor Working 
building floor area l . 85E - 3 
ccs 
- 4:53 Reactor ccs ccs 
building 3.7 0E- 3 
RPVCS 
Electrical Generator 
Reactor 
4:75 busbars & busbars 1 .6 7E-2 
building 
breakers 
Generator 
circuit 
breakers Nwu 
Brake air- :. L.l8fiA.li"t' 
cooled I "-' 
transformer 
56 
Brake circuit 
breaker 
Earthing 
transformer & 
resistor 
Voltage 
transformer 
Excitation 
transformer 
Earthing switch 
Generator 
earthing 
equipment 
Power DC chargers 
electronics 
room 
UPS 
Reactor 
4:71 EMB cabinets, 9 . 25E-3 building 
including test 
equipment 
PLC equipment 
RCS & PEI 
4:73 Reactor RCS room 
1.85E-3 
building 
11:80 Reactor RPV RPV 
building 
RCCS 
7.40E-3 
RSS 
RCS 
57 
3.2.4 Fire Propagation Analysis 
Two stage event trees for two redundant components were used to 
divide the fire model into four elements, Ignition, Detection, 
Suppression and Propagation. It was assumed that once fire 
started in one component it would potentially propagate to the 
next. The loss of all equipments in a room frequency was obtained 
using the Risk Spectrum code. The input data for the failure to 
suppress and detect were from literature [5]. 
The plant was divided into seven subsections and an event tree 
was constructed for each subsection, namely, the cable room, the 
control room, the battery room, the generator room, the 
electrical switchgear room, the auxiliary building, and the 
reactor building. 
58 
Figure 3-1 Event tree for the control room 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
.. 4.lOE-03 (None) 
.... 
D, s 2 
........  1.70E-04 (A) 
p 
- -
S1 s, ......  
l.64E - 09 (A&B) 
-
i p ..... . l.70E-05 (A) Success 
s2 
l ......  2 .04E-05 (A) Failure D2 
p 
-
s 2 
.... 1. 97E - 10 (A&B) 
- .... 
D, 
-
D2 ........  2. 33E- ll (A&B) 
-
p .. 
.... 2.04E-06 (A) 
59 
Figure 3-2 Event tree for the cable room 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
.. 7 . 20E - 03 (None) 
~ 
DI sz 
........  2.98E- 04 (A) 
p 
- -
SI s, .. 
~ 
2. BBE-09 (A&B) 
-
i p ......  2.98E-05 (A) Success 
sz 
l .. ~ 3.59E-05 (A) Failure Dz 
p 
-
sz 
........  3. 4 6E- 10 (A&B) -
DI 
-
Dz .... 
~ 4. 09E - 11 (A&B) 
-
p .. 
~ 3.59E-06 (A) 
60 
Figure 3-3 Event tree for the electrical switchgear room 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
1.30E- 02 (None) 
.~...  
D, sz .. 
~ 5.39E-04 (A) 
p 
-s, -s, .......  
5.20E - 09 (A&B) 
-
i p .~...  5 . 39E-05 (A) Success 
sz 
i .~...  6.47E- 05 (A) Failure Dz 
p 
-
sz .. 6. 25E-10 (A&B) 
- ~ 
D, 
,-- -Dz .......  7 . 38E-11 (A&B) 
$~ 
lf -p .~.. . 6.47E-06 (A) 
61 
Figure 3-4 Event tree for the battery room 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
...... 
2.90E-03 (None) 
. 
D, s 2 .......  1.20E-04 (A) 
p 
- -
S1 S1 .......  
l.16E-09 (A&B) 
-
i p .......  1.20E-05 (A) Success 
s 2 
iF ... ~ l.44E - 05 (A) ailure D2 
p 
-
s 2 
.......  1. 40E-10 (A&B) -
D, 
-
D2 .......  1.65E-11 (A&B) 
-
p ... 
~ 1.44E-06 (A) 
62 
Figure 3-5 Event tree for the generator switchgear room 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
... 1.90E-02 (None) ... 
DI s2 
........  7.87E-04 (A) 
p 
- -
S1 sry .~.. . 
7. 60E-09 (A&B) 
-
i p ......  7.87E-05 (A) Success 
s 2 
i ......  9.46E -05 (A) F ailure D2 
p 
-
S2 
.......  9.14E-10 (A&B) -
DI 
-
D2 ........  1. OSE-10 (A&B) 
-
p 
.~..  9.46E-06 (A) 
63 
Figure 3-6 Event tree for the reactor building 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
.......  S .OOE - 02 (None) 
DI s 2 
.......  2. 07E - 03 (A) 
p 
- -
SI s ? ....... . 
2. OOE - 08 (A&B) 
-
i p .......  2.07E-04 (A) Success 
s2 
l .......  2.49E - 04 (A) Failure D2 
p 
-
s2 
.......  2. 41E-09 (A&B) -
DI 
-
D2 
.......  2 . 84E-10 (A&B) 
-p 
.......  2.49E - 05 (A) 
64 
Figure 3-7 Event tree for the auxiliary building 
Stage 1 Stage 2 Components 
Component A Component B lost 
Ignition I Detection I Suppression Propagation I Detection I Suppression frequency 
... 3.40E-02 (None) ... 
D, s 2 
......  1. 41E-03 (A) 
p 
- -
sl S, ......  
1.36E-08 (A&B) 
-
i p .......  l . 41E-04 (A) Success 
s 2 
l ......  l.69E-04 (A) Failure D2 
p 
-
s 2 
......  1. 64E - 09 (A&B) -
D, 
-
D2 ......  1 . 93E-10 (A&B) 
-
p 
......  1. 69E-05 (A) 
3.2.5 Plant system analysis 
The system analysis has been done for the PBMR plant in t he PBMR 
PRA Report (Chapter 5 Data Analysis ) . For this project only the 
safety s y stems and their support s y stems were considered i n the 
plant system analysis for fire accidents. 
As mentioned earlier in the document, fire is unpredictable, it 
can cause a short circuit instead of an open circuit, fire also 
tends to induce multiple system inter dependencies. It is 
therefore important to note all the support s y stems for different 
components and s y stems of the plant. 
As shown earlier (Table 2 Safety System Components and their 
Support Systems (System Dependencies)) all safety systems have a 
number of components that assist them in functioning. These 
components also hav e support s y stems, which aid them in 
functioning . By Table 2 Safety Sy stem Components and their 
Support Systems (Sy stem Dependencies ) : 
RCS: Some of the RCS components are supported by the RPS and the 
EPS, and in some also the power plant electrical s y stem is 
included. 
RSS: Some components of the RSS are supported by the RPS and the 
EPS, and in others also the power plant electrical s y stem is 
included. 
66 
CCS: Some of the CCS components are supported by the ACS-
auxiliary closed circuit and the ACS-back up cooling tower closed 
circuit, and some are supported by the WMPS , whilst others are 
supported by the EPS, the OCS, and the power plant electr i cal 
system. 
SBS: The components of the SBS are supported by the EPS, OCS, and 
the power plant electrical system. 
RCCS: Some components of the RCCS are supported by the ACS-open 
circuit, and some by the ACS-back up cooling tower closed 
circuit, some by the WMPS, and others by the EPS, OCS, and the 
power plant electrical system. 
Support Systems 
Support systems also have components, which in turn also have 
their support systems. 
RPS : Its components are supported by the power plant electrical 
system. 
EPS: Its components are supported by the power plant electrical 
system. 
OCS: Components are supported by the power plant electrical 
system. 
Power Plant Electrical System : The components are supported by 
the EPS and the OCS . 
67 
ACS-auxiliary closed circuit : Some of this s y stem's components 
are supported by the ACS-open circuit, and some by the WMPS, and 
others by the EPS, OCS, and the power plant electrical s y stem. 
ACS-back up cooling tower closed circuit: Some of its components 
are supported by the WMPS, and others by the EPS, OCS, and the 
power plant electrical system. 
WMPS: Components are supported by the EPS, OCS, and the power 
plant electrical system. 
CHAPTER 4 
RESULTS AND HYPOTHESIS 
This section presents the results of probabilistic fire risk 
assessment performed for the PBMR. The results are presented in 
terms of the fire frequency in the different fire compartments. 
The frequency estimates obtained here are a result of quantifying 
the detection and suppression failure fault and event trees. 
Both the initiating event frequencies of fire and fire event tree 
sequence frequencies are presented. Critical safety equipment and 
their support systems, including cables and power, were 
identified in each compartment. 
Once the sequence frequencies have been calculated, one 
determines whether the fire in a particular compartment could 
result in an internal initiating event. That is, whether the 
fire will cause a pipe break, reactivity excursion, or loss of 
the main power s ystem. If an internal initiating event could be 
caused as a result of fire in a particular room, then the 
frequency is compared to a predetermined value of 10 -2 per year . 
This means that only the contribution of fire to the release of 
radionuclide greater than 1% is considered. If the resulting 
fire sequence frequency is greater than this value , then the 
69 
particular frequency i s retained. This is then multiplied by the 
internal initiating event frequency to determine the overall 
risk. 
4.1 Results 
J 
4 .1.1 Fire Initiating Event Frequencies for the different 
rooms of the PBMR power plant. 
Table 4 .1 below presents the calculated fire initiating event 
frequencies per identified room. These frequencies were 
determined using the methodology presented in Chapter 3 . The 
table provides initiating ev ent frequencies of fire in various 
compartments in the reactor, namely, control room, cable 
spreading room, diesel generator room, switchgear room, battery 
room, auxiliary building, and reactor building. The frequencies 
from Tables were obtained from Nureg 1150 [3] of which the values 
are exactly the same as the values in Table 3.4 since the PBMR 
plant does not have its own statistical data that can be used to 
stipulate the frequencies. For simplicity, it is assumed that 
once the fire starts, it spreads. This means that the mitigation 
system response has failed. Thus, the failure frequency is 1. 
Thus, the total sequences frequency is the same as the initiating 
event frequency of fire. 
70 
Table 4.1 Fire Initiating Event Frequencies 
Fir e Frequency 
Pl ant Location 
(per reac t or year) 
Control room 4 . lE- 03 
Cable spreading room 7.2E-03 
Diesel Generator room 1.9E-02 
Switchgear room 1.3E-02 
Battery room 2.9E - 03 
Auxi liary b u i ldi ng 
3 . 4E- 02 
Reactor building 
7.2E-03 
From the table, only three rooms may potentially contribu te to 
internal ev ents and these are diesel generator room, switchgear 
room and auxiliary building. However, loss of equipment in these 
rooms will not lead to any radiologi cal release. However , this 
is a point, which requires further studies to determine if there 
is any release as a result of fire. 
71 
4.2 Hypothesis 
4.2.1 Fire occurrence frequencies 
The occurrence of fire in the PBMR power plant is not expected to 
contribute significantly to the total risk to the public and 
plant personnel. This is a result of the inherent safety of the 
plant . Also, The fire frequencies for the Pebble Bed Modular 
Reactor plant are expected to be much less than the ones from the 
Light Water Reactors, Pressurized Water Reactors, and the Boiling 
Water Reactors, once it has gathered its own empirical data, 
because of its inherent safety. Another factor that may add to 
the fire frequencies being low is the unique design of the plant 
in which very small numbers of combustible materials are used 
where their use could not be avoided; the cable routings take 
different directions so that at least one cable route is 
available if the other one is knocked out by fire during a fire 
accident; all major pumps are surrounded by fire barriers to 
minimize the spread of fire from one channel to the next during 
fire accidents. 
4.2.2 Loss of equipment frequencies (Fire Propagation) 
The obtained frequencies seem to follow a similar pattern for all 
the different rooms in the plant. According to these results it 
seems highly unlikely that fire will propagate from one equipment 
to the next in a room, this is shown by v ery low frequencies 
(close to impossible ) that a fire accident may destroy more than 
72 
one system in a room without being noticed or without something 
being done . 
When using the Risk Spectrum code it was noticed that as the time 
between detector tests (during maintenance of the detectors ) 
increased, the loss of equipment frequencies also increased but, 
the more frequent the tests are done the lower the loss of 
equipment frequencies become (close to impossible ). The 
disadvantage however of more frequent tests is that it i s very 
expensive. 
CHAPTER 5 
SUMMARY, CONCLUSION, AND RECOMMENDATIONS 
5.1 Summary 
In this report the Probabilistic Risk Assessment methodology was 
used to assess the frequencies of occurrence of fire. To do 
this, the safety functions of the plant were determined . Then 
the safety systems that would be needed to perform these safety 
functions were identified together with their support systems. 
The safety functions considered are: control of heat generation, 
heat removal . 
To determine the frequency of fire occurrence, areas were 
identified where the various safety system are located. The 
plant's building layout was used for this purpose. The various 
areas were further broken down into smaller compartments for ease 
of representation. Since there is no operational history for the 
PBMR, data from the Light Water Reactors, Pressurized Water 
Reactors, and Boiling Water Reactors was used to derive the fire 
occurrence frequencies for the different locations in the plant. 
To obtain frequencies for each compartment in the PBMR plant, the 
frequencies for different areas from the LWR and PWR plants were 
74 
broken down according to the number of systems present in each 
compartment in the PBMR plant . 
Sequence frequencies are generally determined by combining the 
initiating ev ent frequencies with the failure frequenc i es of 
various systems considered in the fire mitigation systems. This 
part models the propagation of fire from one system to another 
prior to detection and suppression failures in a room, and the 
resulting loss of equipment . To simplify the analysis, the 
conditional probability of the sequence was assumed to be unity . 
The total fire frequency was then the equivalent to the 
initiating event frequency . 
The Fire PRA analysis was done to investigate the impact of fire 
on the safety of the PBMR plant and also on the existing PBMR PRA 
risk frequency . This analysis was also done to determine whether 
the PBMR plant would be able to reach a safe reactor shut down in 
case of a fire accident , and most importantly to see if the total 
risk due to the operation of the plant is still within t h e risk 
limit specified by the National Nuclear Regulator . 
5.2 Conc lusion 
From the frequencies presented in Table 4 .1, it is clear that 
fire is not expected to contribute significantly to the overal l 
risk of the PBMR . The initiating event frequencies for the 
various locations were below a predetermined cut-off frequency . 
75 
Thus, the impact of fire on the existing PBMR PRA risk fr equency 
is negligibly s mall . 
5.3 Recommenda t ions 
This study was a preliminary assessment of the impact of fire in 
the PBMR . The study was limited as a result of the fact that the 
data used as input was generic. Therefore, when new and relevant 
operational data becomes available, the study will have to b e 
updated . Also, this study did not involve sensitivi t y and 
uncertainty analysis. This is essential in any PRA study . 
Future studies should include these assessments. 
Glossary of terms 
TERM DEFINITION 
An unplanned event or sequence 
of events that results in 
undesirable consequences. An 
incident with specific safety 
Abnormal Condition consequences or impacts such as 
damage to one or several 
barriers, thus leading to the 
release of radioactive 
material . 
A system that has specifically 
Active Cooling System (ACS) designed controlled components 
to cool a system or component 
A system to provide compressed 
air of suitable quality to : 
Supply maintenance outlets. 
Compressed Air System (CAS) Supply pneumatic va l ve 
actuators. 
Pressurize header tanks of the 
ACS. 
A system for the removal of 
decay heat when the Brayton 
cycle is not operating. It also 
Core Conditioning System (CCS) provides for helium flow 
through the reactor core 
heat -up during start-up 
operations 
It provides heat removal at 
pressures lower than the main 
Cooling Water (CW) 
operating system pressure at 
all times during operation. 
A condition that deviates from 
Event 
normal operation 
A specific unplanned series of 
events composed of an 
initiating event and 
Event Sequence 
intermediate events that may 
lead to a deviation from normal 
operation 
External conditions that cause 
External Events deviations from normal 
operation . The examples are: 
77 
Floods, earthquakes, v o l cano, 
terrorism, fires etc. 
A qualitativ e method of s y stem 
analysis which inv olv es the 
study of the failure modes that 
Failure Modes and Effects can exist in every component of 
Analy sis (FMEA) the system, and the 
determination of the causes and 
of the effects of each failure 
mode 
A variation of FMEA tha t 
includes a quantitative 
Failure Modes, Effects and 
estimate of the significance of 
Criticality Analy sis (FMECA ) 
the consequence of a fai l ure 
mode 
The termination of the ability 
Failure of an entity to perform a 
required function. 
The effect by which a fa ilure 
Failure Mode 
is observed 
A system that provides early 
warning of fire in the module, 
PPSB and Auxiliary building, 
Fire Protection Sy stem (FPS ) 
suppresses fire, is used for 
fire extinguishing purposes and 
prev ents the spread of fire. 
A s y stem for loading and 
Fuel Handling and Storage 
removal of fuel and graphite 
Sy stem (FHSS ) 
spheres from the reactor 
A s y stem that maintains 
specified environmental 
parameters of temperature and 
(where required) humidity and 
Heating, Ventilation and Air- minimizes the internal building 
conditioning (HVAC) System contamination b y air renewal , 
maintaining direction of 
airflow and removing 
radioactive particles b y 
purging and filtering. 
A s y stem for 
Helium regulating the Inventory Control Sy stem 
quantity and quality of helium 
(HICS ) 
in the PCU. 
The first event in an ev ent 
Initiating Ev ent 
sequence. 
Internal Events Internal conditions that cause 
dev iations from normal 
78 
operation, and the examples 
are : loss of coolant, 
small/medium/large pipe break, 
control rod failure, accidental 
drop of the SAS etc . 
The Main Power System 
physically combines the 
functions of the Reactor Unit 
Main Power System (MPS) 
(RU) and Power Conversion Unit 
(PCU) to convert nuclear energy 
into electricity 
The ability of an item under 
given conditions of use, to be 
restored to a state in which it 
can perform its intended 
Maintainability 
function, when maintenance is 
performed under given 
conditions and using stated 
procedures and resources. 
A single, stand-alone PBMR, 
Module containing all systems for 
operation. 
A system that controls and 
monitors major plant, aux iliary 
systems and support systems. 
Start-up, power generation and 
Operational Control System 
shutdown are controlled through 
(OCS) 
the OCS. The OCS also stores 
process data and makes data 
available for other external 
systems 
A high-temperature, direct 
Pebble Bed Modular Reactor cycle, helium-cooled nuclear 
(PBMR) reactor with fully ceramic 
spherical fuel elements. 
Manoeuvring or changes in 
Plant Mode process variables within a 
specific state 
Plant conditions or parameters 
that are fundamentally 
different from each other. The 
Plant State 
plant can continue 
'indefinitely' in any specific 
state 
Changing from one plant state 
Plant Transition to another. The plant can only 
be in transition for a specific 
79 
time period, and has defined 
initial and end conditions. 
An assembly of equipment, 
comprising the high-and low-
pressure turbo-units, power 
turbine generator, a 
Power Conversion Unit (PCU) 
recuperator and coolers, 
combining to convert the heat 
from the helium into 
electricity 
An assembly to convert fluidic 
Power Turbine Generator (PTG) 
helium energy into electricity 
The system responsible for 
controlling the reactor during 
normal operation, effecting a 
Reactivity Control and Shutdown fast transition to a sub-
System (RCSS) critical state on demand, and 
maintaining the reactor sub-
critical indefinitely in a cold 
condition (100 °c) 
A system to transfer the heat 
from the reactor cavity during 
Reactor Cavity Cooling System all modes of reactor operation 
(RCCS) to the ultimate heat sink (sea 
or to the atmosphere, when the 
sea is not available). 
The vessel, which supports the 
reactor core structures and 
Reactor Pressure Vessel (RPV) 
forms part of the MPS pressure 
boundary. 
A system for maintaining the 
RPV at a homogeneous 
Reactor Pressure Vessel temperature, in order to 
Conditioning System (RPVCS) maintain operating temperatures 
in the range for which the 
materials have been qualified. 
A system that shuts the reactor 
down when certain values as 
specified in the General 
Operating Rules (GOR) and/or 
Reactor Protection System (RPS) technical specifications have 
been exceeded . The RPS also 
prevents restart of the reactor 
until normal operating 
conditions have been restored. 
Recuperator A device for returning heat in 
the gas downstream of the 
I WU- I 
9 IR RARV 
80 
turbines to the gas returning 
to the reactor, thus improving 
cycle efficiency. 
Fuel that has reached its 
Spent Fuel 
maximum design burn-up. 
A system that circulates helium 
gas through the main gas 
circuit of the MPS in order to: 
Start the Brayton thermodynamic 
Start-up Blower System (SBS) 
cycle or 
Cool the core intentionally if 
the Brayton cycle is not 
operational 
The control room in the 
Power Pant Control Room Services Building from where 
all modules will be controlled. 
Fuel that has undergone a 
partial fission reaction, but 
Used Fuel 
has not yet reached its maximum 
design burn-up. 
81 
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