Passive Safety Injection System Design and Simulation for Small Scale Pressurized Water Reactor

Muhammad Tahir

A dissertation submitted for the degree of “Doctor of Philosophy” (PhD) in

Pakistan Institute of Engineering and Applied Sciences Islamabad Pakistan May 2011 Declaration

I declare that all material in this thesis which is not my own work has been identified and that no material has previously been submitted and approved for the award of a degree in this or in any other university.

Signature: ______

Author’s Name: (Muhammad Tahir)

Supervisor

Dr. Imran Rafiq Chughtai

Principal Engineer Department of Chemical and Materials Engineering Pakistan Institute of Engineering and Applied Sciences [PIEAS] Islamabad, Pakistan.

Head, DNE, PIEAS

Acknowledgements

All praises and thanks to God, the most Merciful, Compassionate, Gracious and Beneficent who has created man and is a source of knowledge and wisdom.

At the very outset, I am thankful to my supervisors, Dr. Imran Rafiq Chughtai, PE and Dr. Muhammad Aslam, CE for their supervision, technical advices throughout the investigations and preparation of this manuscript. I am greatly indebted to director Imtiaz Rabbani for his administrative guidance.

I am also grateful to my professors at PIEAS especially Dr. Muhammad Aslam, Dr. Tehsin Hamid, Dr. Naseem Irfan, Dr. Mansoor Hamid Inayat, Dr. Nasir Majid Mirza, Dr. Sikandar Majid Mirza and Dr. Muhammad Tufail.

I am thankful to Dr. Muhammad Arfin Khan for his assistance in visiting Texas Tech University USA and guidance in educational and research activities.

I am also grateful to my friends and colleagues who ensured a creative and good working environment and helped me in technical and non-technical matters. My special thanks are due to Dr. Munawar Iqbal, and Dr. Basharat Ali. I am also thankful to HEC for grants in favor of foreign visits.

I am also thankful to my wife, daughters and sons who have been missing me during my long working hours at PIEAS. I would like to express my dearest feelings towards my parents for their special prayers. At the end, I am grateful to all those who have always wished to see me glittering high on the skies of success, may ALLAH bless them with good health and long lives!

Muhammad Tahir

PhD research work publications

International Publications M. Tahir, I. R. Chughtai and M. Aslam, Qualitative response of SIS for a large size break LOCA in the SRC cold leg, Annals of Nuclear Energy 34(2007) 922- 925.

M. Tahir, I. R. Chughtai and M. Aslam, Response of Proposed Passive SIS for an Intermediate Size Break on CHASNUPP-1, Annals of Nuclear Energy 35(2008) 1986-1993.

Conference Papers M. Tahir, et al, Proposed Passive SIS for Chasnupp-1 Type Reactors, International Workshop on Passive Safety Systems in Advanced PWRs, 28-30 April 2008, Shanghai Jiao Tong University, Shanghai, China.

M. Tahir, et al, Comparative Study of the Passive Safety and the SIS under Loss of Coolant Accident, paper presented in International Conference Spring 2009 Meeting of the Texas Sections of the APS, AAPT, and SPS, 02-04 April 2009, Tarleton State University ,Stephenville, TX, USA.

M. Tahir, et al, Simulation of a Passive Safety Injection System for a Small Scale Advance Pressurized Water Reactor Design, paper presented in international conference, Modern trends in physics research, MTPR-10, 12-16 December 2010, Cairo University, Cairo, Egypt.

Table of contents

LIST OF FIGURES ...... 7

LIST OF TABLES ...... 10

CHAPTER 1 ...... 17

1 INTRODUCTION ...... 17

1.1 PROBLEM DEFINITION ...... 19

1.2 RESEARCH OBJECTIVES ...... 20

1.3 THESIS ORGANIZATION...... 21

CHAPTER 2 ...... 24

2 LITERATURE SURVEY ...... 24

2.1 REVIEW OF PWR TECHNOLOGY ...... 24 2.1.1 NPP safety and safety systems ...... 25 2.1.2 Active and passive safety systems ...... 26 2.1.3 Passive safety systems ...... 27 2.1.4 Applications of passive safety systems in large size reactors ...... 28 2.1.5 Hybrid passive safety systems ...... 31

CHAPTER 3 ...... 33

3 MATHEMATICAL MODELING ...... 33

3.1 THREE EQUATION MODEL ...... 34 3.1.1 Mass balance equation ...... 34 3.1.2 Momentum balance equation ...... 35 3.1.3 Energy balance equation ...... 36

3.2 FIVE EQUATION MODEL ...... 36 3.2.1 Mass balance equations ...... 37 3.2.2 Momentum balance equation ...... 37 3.2.3 Energy balance equation ...... 38

3.3 THERMO-HYDRAULIC CORRELATIONS ...... 39

3.4 DECAY HEAT ...... 40

3.4.1 Fission heat after shutdown ...... 41 3.4.2 Fission product decay Heat ...... 42

3.5 ADVANCE PROCESS SIMULATION ENVIRONMENT ...... 43

CHAPTER 4 ...... 47

4 NUCLEAR SAFETY SYSTEMS ...... 47

4.1 EMERGENCY CORE COOLING SYSTEM ...... 47 4.1.1 Charging system ...... 48 4.1.2 Boron injection system ...... 49 4.1.3 Safety injection system ...... 49 4.1.4 Accumulator injection system ...... 50 4.1.5 Residual heat removal system ...... 51 4.1.6 Role of ECCS during various LOCAs ...... 53

4.2 SIMULATION OF EMERGENCY CORE COOLING SYSTEM ...... 56 4.2.1 Simulation of reactor trip system ...... 57

CHAPTER 5 ...... 71

5 PROPOSED PASSIVE SAFETY SYSTEM ...... 71

5.1 THERMAL HYDRAULIC DESIGN OF THE PROPOSED SYSTEM ...... 71

5.2 CONTROLS IMPLEMENTED ...... 76

CHAPTER 6 ...... 78

6 PROPOSED SYSTEM APPLICATIONS ...... 78

6.1 VALIDATION OF REFERENCE POWER PLANT MODEL ...... 78

6.2 PROPOSED PASSIVE SAFETY SYSTEM RESPONSE ...... 78 6.2.1 Intermediate break LOCA ...... 79 6.2.2 Proposed system‟s response for large break ...... 94

CHAPTER 7 ...... 107

7 CONCLUSION AND FUTURE RECOMMENDATIONS ...... 107

REFERENCES ...... 110

APPENDIX ...... 114

Reference power plant simulation ...... 114

List of Figures

Figure 2-1 AP600/AP1000 passive safety systems ...... 30

Figure 2-2 AP600/AP1000 containment cooling system ...... 30

Figure 4-1 Block diagram of charging and boron injection system ...... 48

Figure 4-2 Block diagram of safety injection system ...... 50

Figure 4-3 Block diagram of accumulator injection system ...... 51

Figure 4-4 Block diagram of residual heat removal system ...... 52

Figure 4-5 Block diagram of emergency core cooling system ...... 53

Figure 4-6 Safety injection system simulation diagram ...... 62

Figure 4-7 Accumulator injection system simulation diagram ...... 65

Figure 4-8 Residual heat removal system simulation diagram...... 67

Figure 5-1 Proposed Passive Safety Injection System ...... 73

Figure 5-2 Proposed Passive Injection System Loop-A ...... 74

Figure 5-3 Proposed Passive Injection System Loop-B ...... 74

Figure 5-4 Proposed system delivery in reactor coolant system ...... 75

Figure 5-5 Control implementation for loop-A ...... 77

Figure 5-6 Control implementation for loop-B...... 77

Figure 6-1 Reactor coolant system showing LOCA-valve ...... 80

Figure 6-2 Proposed system nodalization diagram ...... 81

Figure 6-3 Flow through break with passage of time ...... 84

Figure 6-4 Variation of pressurizer liquid level with time ...... 85

Figure 6-5 PSIS, accumulator and RHR flows with time ...... 86

Figure 6-6 CMT-B liquid level with time ...... 87

Figure 6-7 Variation in accumulator liquid level with time ...... 88

Figure 6-8 Normalized reactor power ...... 89

Figure 6-9 Charging system response ...... 90

Figure 6-10 Loop average and fuel average temperature ...... 91

Figure 6-11 Variation in fuel enthalpy with time ...... 92

Figure 6-12 Maximum fuel and clad temperatures ...... 93

Figure 6-13 Total make-up flow rate ...... 94

Figure 6-14 Normalized reactor power ...... 99

Figure 6-15 Flow through simulated break ...... 99

Figure 6-16 Pressurizer liquid level ...... 100

Figure 6-17 Charging pumps A, B flow-rates ...... 100

Figure 6-18 PSIS, accumulator, RHR flows and cold-leg pressure ...... 101

Figure 6-19 Response of CMT-B liquid level with time ...... 102

Figure 6-20 Response of accumulator liquid level with time ...... 102

Figure 6-21 Variation in loop average and maximum fuel temperature ...... 103

Figure 6-22 Accumulator flows comparison for a spectrum of break sizes .. 104

Figure 6-23 Accumulator flow comparison for different break sizes ...... 104

Figure 6-24 Residual heat removal system flow for spectrum of breaks ..... 105

Figure 6-25 Comparison of cumulative flows for different break sizes ...... 105

Figure 6-26 Maximum cladding temperature for different break sizes ...... 106

List of Tables

Table 2-1 LOCA classification ...... 26

Table 4-1 Safety injection signal generation parameters ...... 57

Table 4-2 A list of reference power plant trips...... 58

Table 4-3 Operating modes of reference power plant ...... 60

Table 4-4 List of important variables represented in simulation ...... 69

Table 5-1 Design parameters of proposed system tanks ...... 75

Table 5-2 Design parameters of proposed system isolation valves ...... 76

Table 5-3 Design parameters of proposed system check valves ...... 76

Table 6-1 Steady state safety parameters values for reference power plant . 79

Table 6-2 Chronological sequence of events ...... 83

Table 6-3 Reference power plant parameters ...... 95

Table 6-4 Chronological sequence of events ...... 96

Table 6-5 Passive safety system and safety injection system comparison..... 96

Abstract

Safety and availability are prime factors for nuclear power plant operation. Safe operation requires a well-built backup of safety systems for saving plant capital cost, environment and the public. The backup system is maintained with the use of active and passive safety systems in the form of engineered safety features. Traditionally, active safety systems have been utilized for counteracting accidental conditions. These systems require proper and timely operator actions, which is some time misleading. Now a day, passive safety systems are becoming more popular due to their dependence on forces of nature for operation and actuation. Passive operations include, under gravity flow, natural recirculation of fluid and nitrogen gas pressure. For such type of systems, when certain conditions are met and the passive operation is started automatically. Therefore, world nuclear community has started using passive safety systems in the present nuclear power plant technology due to the simplicity in operation, maintenance and safety enhancement.

This thesis concentrates on the augmentation of passive safety features in small scale pressurized water reactor design. The research was started with the study and simulation of a small scale reference power plant. The scope of simulation includes safety systems including necessary nuclear and conventional island. The individual process systems and related electrical systems are simulated and integrated within the frame work of their respective instrumentation and control to form a standalone simulated model for a reference power plant. Using this model, design basis accident has been modeled and the response of the safety systems together with related primary systems has been observed. Satisfactory results have been experienced in this regard.

The research was extended by designing and simulating a passive safety injection system. This proposed system consists of many passive components and functions in place of an existing safety injection system for mitigating loss of coolant accidental condition. The use of proposed passive system has been suggested only for intermediate type pipe breaks because for small breaks, the depressurization is slow and only high pressure charging system is utilized whereas for larger breaks, the depressurization is very fast and low pressure safety injection system is actuated rapidly. Therefore, in a simulation test run this proposed system has been tested and verified for intermediate coolant loss accident. It has been found that the response of the proposed passive system is satisfactory and it keeps all necessary safety parameters within range. Through this research, it is concluded that the proposed passive system could be a potential candidate for handling intermediate type breaks representing loss of coolant accidents in small scale pressurized water reactors. For other accidental conditions of the plant like steam generator tubes rupture and steam line break, the use of conventional way of accidental management has been suggested.

Keywords: Advance NPP, Passive safety, Analysis, LOCA, Reliability

Nomenclature

Acronyms and abbreviations

APROS advance process simulation environment

PWR pressurized water reactor

CMT-A core makeup tank-A

CMT-B core makeup tank-B

ACC accumulator tank

LOCA loss of coolant accident

NPP nuclear power plant

RWST refueling water storage tank

RHR residual heat removal system

ECCS emergency core cooling system

MPa mega pascal

RELAP computer code, reactor leak and analysis program

Loop-A Reactor coolant system loop A

Loop-B Reactor coolant system loop B

I & C instrumentation and control

SRC reactor coolant system

SCV chemical and volume control system

SCS containment spray system

SIS safety injection system

PSIS passive safety injection system

SBLOCA small break LOCA

TMI Three Mile Island

ADS automatic depressurization system

RPP reference power plant

PACTEL parallel channel test loop

MIT Massachusetts Institute of Technology ppm Parts per million

Symbols

ρm Mixture density

vm Mixture velocity

ρv Vapor density

ρl Liquid density

VV vapor volume

α, 훼푣 void fraction

′′ qx Heat flux in x direction

Gm Mixture mass flux z Distance along z-axis p Fluid pressure g Acceleration due to gravity

Fwm Mixture wall friction

fm Component momentum loss

hm Mixture enthalpy

qwm Wall heat transfer

m l , m v Liquid mass flow, gas mass flow

′′′ q0 Volumetric heat generation rate

qcr Critical heat flux

φ0 Steady-state neutron flux

β Delayed neutron fraction

ρ Reactivity l Prompt neutron life time

γ1 Decay constant for longest lived delayed neutron precursors

Q Core power after shutdown

Q 0 Steady state core power

Pβ Power due to β rays

Pγ Power due to γ rays

P0 Steady- state power

Re Reynolds number

Pr Prental number h Heat transfer coefficient

D Tube diameter k Thermal conductivity

Chapter 1

1 Introduction

Nuclear power plant is a tremendous source of energy and plays a vital role in providing sustainable, cheaper and clean energy [1]. Safety and availability are prime factors for any nuclear power plant operation. Safety is related to the control and confinement of radioactivity inside the reactor core all the time. In traditional design, to confine radioactivity inside reactor core there exist seven safety barriers namely: reactor fuel, fuel cladding, reactor coolant, reactor pressure vessel, reactor containment, remote site and evacuation plane [2]. It is necessary to protect and strengthen these safety barriers during all operating modes. In case of an abnormal condition, radioactivity may breach these safety barriers and can go into environment air or water hazardous for public health and safety.

The greater risk to the reactor core and safety barriers is from Loss of Coolant Accident (LOCA). In this abnormal condition, the coolant performing the function of heat removal from the reactor core is leaked out from the reactor coolant system pressure boundary and function of core cooling is degraded. The reactor coolant system inventory is decreased and the reactor is shut down by reactor protection system. After shut down of the reactor, the rise in temperature of core is due to presence of decay heat [3]. If the availability of cooling water is not restored well in time, the decay heat could increase the temperature of the core to such an extent where fuel could be damaged. Such type of phenomenon has happened in Three Mile Island and Chernobyl

accidents[4, 5]. The latter was ended with causalities due to high amount of dose to a large number of peoples.

Now a days, more and more new reactors are being built around the globe to meet growing energy requirements. Due to unavailability of space, some of these reactors would be proposed to be built near population zones or the population around nuclear power plant would increase by the passage of time. The importance of safety for these nuclear power plants has increased manifold due to spreading of population zones and the safety requirements of such reactors would be different and more stringent than the existing nuclear power plants. For such reactors, evacuation plan and remote site planning will not work properly. The nuclear accidents would not be affordable in future reactors under aforementioned conditions.

In the typical nuclear power plant design, Emergency Core Cooling System (ECCS) is employed to cool and protect the reactor core under LOCA conditions. This system injects borated water into the reactor core, drives and maintains the core in a safe state. Emergency core cooling system consists of redundant cooling trains and active components like pumps and valves which require external intelligent signals for their operation. These active systems employ redundant power sources for their operation and complicate the design and operation of the overall plant, particularly the safety systems. Moreover, correct operator actions are required to operate these systems in a particular situation for mitigating an accidental condition. The accident of Three Mile Island was made severe due to operator‟s actions by shutting down its safety systems which were designed for such accidents.

In order to overcome the above mentioned problems with active safety systems and to further ensure the reactor safety, passive safety systems are being introduced in the modern nuclear power plant designs worldwide. Instead of redundant external power sources for active components, the passive components utilize natural forces e.g. evaporation, natural circulation,

compressed gas and gravity for their actuation. A passive system [6] is composed of passive components and utilizes active components in a very limited way to start the passive operation thus eliminating the operator requirements. Some examples of these passive systems reported in the literature are gravity drain tanks, elevated tank, natural circulation loops (core make-up tanks), pre-pressurized core flooding tanks, passive heat exchangers and sump natural circulation [7]. When thermal hydraulic conditions exceed predefined limits the passive operation is started automatically and is more reliable as compared to active operation [8]. The active safety systems require redundant electric power supplies with multiple active components, whereas, passive systems offer simplicity in installation, operation and maintenance, with reduced cost with improved safety [9].

The nuclear power plants can be categorized based upon their capacity as small, medium and large. Although, analyses of passive safety systems for medium to large size reactors have been in focus of recent research[10], however, small scale reactors have not been analyzed in this context. In the perspective of power planning in Pakistan, the importance of small scale nuclear power plants cannot be over emphasized.

1.1 Problem definition

The idea of replacing an active safety system with a passive one in future designs must, at least theoretically, be exhaustively evaluated from operational as well as safety point of view. This evaluation, however, may require rigorous analysis of the overall plant behavior since various systems like reactor core, main reactor coolant system, decay heat removal system and safety injection system etc, all interact with each other dynamically. This interaction is not only related to thermal hydraulic but also involves control and electrical system considerations. It is very obvious that such system level design proposals can only be analyzed using computer simulation techniques.

Hence, the complex systems of a nuclear power plant require the use of comprehensive computer code that can handle reactor core neutronics, electrical systems, thermal hydraulics and instrumentation & control in a coupled and integrated manner.

The focus of research in thesis will be to evaluate the idea of implementing a safety injection system with passive components in place of an existing active system design. This requires a comprehensive simulation of proposed system including all interacting systems of reference power plant which is assumed to be a typical 325 MWe pressurized water reactor, similar in designs already installed in Pakistan. The behavior of this system must be analyzed for normal operating conditions as well as various accidental scenarios during loss of coolant accidents.

1.2 Research objectives

The specific objective of this research is to design and model a passive safety system for a reference power plant. To accomplish this task, a point wise description of research objectives is given below:

 To model reference power plant with and without the proposed passive safety injection system including all major interacting systems, e.g.:

 Nuclear Island

i. Reactor coolant system

ii. Charging system

iii. Safety injection system

iv. Accumulator injection system.

v. Residual heat removal system

vi. Component cooling water system

 Conventional Island

i. Feed water system

ii. Auxiliary water system

iii. Main steam system

iv. Turbine system

 To propose and simulate a passive safety injection system.

 To simulate a hypothetical accidental condition for observing the response of the active safety injection system and the proposed passive safety injection system for a range of break sizes.

 To make a comparison of the two approaches for safety augmentation.

1.3 Thesis organization

Chapter 1 - Introduction

This chapter deals with the general introduction of nuclear power plant and the importance of safety. The use of active and passive safety systems for handling LOCA has been presented. A brief review of the present status of passive systems under forces of nature, along with an indication towards the areas for further research and main objectives of the research have been defined. Finally an organization chart of the thesis has been given.

Chapter 2 – Literature review

In order to have a complete picture of the present issue related to research problem, literature survey is necessary. The current status of the problem of safety with reference to active and passive has been discussed in this chapter. The main emphasis has been on the safety systems involving passive objects. A brief review of reference power plant safety is also presented in this chapter.

Chapter 3 – Mathematical modeling

The complete modeling of nuclear power plant requires four different areas of reactor simulation. It includes thermal hydraulics, core neutronics, instrumentation & control and electrical system. For finding the consolidated solution of such problems we have used advance process simulation environment. This is modular software and finds its application in simulator and plant analyzer. For finding the thermal hydraulic response of the primary and auxiliary systems fundamental equations of mass, energy and momentum equations have been described in this chapter.

In order to use software for analysis purposes, its validation with experimental data or with some authentic code is necessary. The software we have used has been validated both theoretically & experimentally, the analyzers for both LOVISA and COLA nuclear power plants have been developed using this software, its detail has been provided in this chapter.

Chapter 4 – Nuclear safety systems

This chapter deals with the already existing safety systems of nuclear power plant. Normally operating charging system, safety injection system, accumulator injection system and residual heat removal systems have been discussed. Moreover, the role of these systems under design basis accidents is also described.

Chapter 5 – Proposed passive injection systems

Passive safety is a hot topic of the day, for this purpose and to make a reactor safer, a passive safety injection system is proposed which is closest to the concept of pre-pressurized tanks. The design, simulation and operation of this system has been described in this chapter.

Chapter 6 – Proposed system applications

The proposed passive safety injection system has been put into operation for its performance under design basis accidents. For this purpose this system has been tested for intermediate to larger break LOCAs. The detail of these transients has been provided in this chapter.

Chapter 7 – Conclusion and future recommendations

In this chapter conclusion of research has been provided. The work done so far is in the direction of implementing passive safety features in the present design and thereupon simplifying the nuclear power plant. Simplification makes a process easy to understand and gives a control over the process. Moreover, in order to have completed design of small scale Passive Power Plant, recommendations regarding different passive features have been presented.

Appendix

Systems necessary for simulation of reference power plant systems, have been described in this section. For this purpose, the simulation of reactor coolant system, simulation of chemical and volume control system and simulation of safety systems has been described. Secondly, the data for simulating the auxiliary systems particularly safety injection system, decay heat removal system and accumulator injection systems have been presented. At the last views of published papers are placed.

Chapter 2

2 Literature survey

2.1 Review of PWR technology

The first pressurized water reactor was conceived, designed and commissioned in America on December 18, 1957 [11]. Currently, around 74 percent of all light water reactors in the world are pressurized water reactors and the majority of these are based on Westinghouse pressurized water reactor technology [12]. These types of reactors utilize enriched uranium as fuel and light water as coolant and moderator. A typical pressurized water reactor has the following advantages:

 These reactors are highly stable due to their negative fuel temperature coefficient of reactivity (Doppler Effect) and negative moderator temperature coefficient of reactivity. According to which the power of these reactors decreases as the temperature increases which makes the control of the reactor easy.

 The turbine cycle loop is kept separate from the primary cycle loop, so that the cooling water in the secondary loop may not be contaminated by the radioactive materials present in the primary side.

 These reactors have the ability to passively shutdown the reactor by dropping the control rods under gravity which is required in accidental cases.

2.1.1 NPP safety and safety systems

One of the advantages of pressurized water reactor technology is that the nuclear island is kept separate from conventional island. The special attention is focused on the primary side of the reactor to protect its radioactive boundary. This primary circuit containing reactor core is hazardous for public health and safety if radioactivity is not controlled within safe limits. Multiple barriers around nuclear core are affected badly, if the temperature of the core is not controlled properly. To protect a reactor core, safety barriers are kept within safe limits during all of its operating modes as mentioned in Table 4-3. However, in accidental conditions safety parameters may go beyond safe limits and radioactivity may enter into environment, air or water hazardous for public health and safety.

As mentioned already, for such type of reactors, the greater risk to the reactor core is from LOCA. The cooling water, providing the services of core cooling leaks out from the reactor coolant system piping. It is worth mentioning here that the rate of coolant loss depends upon the break size. On this basis, the LOCA is divided into four main categories namely: large break, intermediate break, small break and very small break. Each type of LOCA has a proper definition [13] depending upon break size in the main coolant system as given in Table 2-1.

In the present pressurized water reactor design, this accidental condition has been handled with the use of an important system called emergency core cooling system [14]. Under accidental conditions this system protects the reactor core by injecting borated water at various values of main coolant system pressures.

Table 2-1 LOCA classification [13] LOCA category Break size Large breaks Equivalent diameter of break larger than 243 mm up to double ended rupture of largest pipe Intermediate break Equivalent diameter lying between 243 mm to 25 mm Small break Equivalent diameter lying between 25 mm to 9.5 mm Very small break Equivalent diameter less than 9.5 mm

The main function of emergency core cooling system is to protect a reactor under accidental conditions by removing decay heat from core. This system performs its role efficiently by using different safety systems of the plant for their safety functions. These systems include: charging system, boron injection system, safety injection system, residual heat removal system and accumulator injection system[15]. These systems perform their role under different thermal hydraulic conditions of the main coolant system. The detail of these systems and their role under different LOCAs has been presented in section 4.1 and section 4.1.6 respectively. The safety system performs differently for different break sizes. The spectrum of breaks in the main coolant loops are classified in Table 2-1.

2.1.2 Active and passive safety systems

In the present pressurized water reactor design, the majorities of systems working for emergency core cooling system are active by design and require external signals for their activation and depends upon redundant power sources for their operation. Moreover timely and correct operator actions are required to use these systems efficiently during an accidental condition. Sometime, operator misunderstands the accidental condition and takes actions which can complicate an accidental condition. Therefore, the need of an operator independent safety system has been acknowledged in the context of nuclear power plant.

A passive system on the other hand utilizes forces of nature for its operation and does not require operator‟s action for its actuation. It starts automatically when certain conditions are met. Therefore, passive system is believed to be more realistic and reliable as compared to active system. A passive system is composed of passive components and uses active components in a very limited way to start a passive operation [5].

To achieve higher safety standard, passive systems relying on gravity, compressed gas, natural circulation and evaporation. It eliminates the operator requirements to start or stop an operation. The utilization of passive safety systems reduces the costs associated with installation, operation and maintenance of active systems that require various pumps with independent and redundant sources of electric power supply systems. Whereas passive systems have simplicity in installation, operation, maintenance, reduced cost with improved safety[7]. Some of these passive systems have been proposed by IAEA whose details are given below.

2.1.3 Passive safety systems

Some passive safety systems have been reported by IAEA for removing decay heat from the reactor core subsequent to a reactor trip after LOCA [7]. A brief summary of these systems is given below:

 Pressurized core flooding tanks

 Elevated tank natural circulation loops

 Under gravity drain tanks

 Passively cooled steam generator natural circulation loops

 Passive decay heat exchangers

 Passively cooled core and sump natural circulation.

These passive systems can be incorporated in any pressurized water reactor with due modifications with good impact on plant safety, economics and simplicity.

2.1.4 Applications of passive safety systems in large size reactors

In this section, we will discuss some of the large size nuclear power plants having adopted or planning to adopt passive safety systems in near future. The first nuclear power plant of this type is AP600 [12]. Next two reactors that are planning to adopt these passive systems are AP1000 and EP1000 [16, 17]. Both units have attempted to use passive safety system extensively for safety enhancement, reduction in cost, simplicity in operation & maintenance.

Advance passive AP1000 is a two loops pressurized water reactor and is updated type of AP600. It has adopted already verified technology spreading over 35 years of working pressurized water reactor experience. The AP1000 and AP600 are the only nuclear power plants utilizing passive safety technology qualified around globe. The key success of passive safety systems is to provide significant development in plant safety, reliability, plant simplification, plant costs and investment protection. The safety of AP1000 has been confirmed by wide-ranging testing, safety investigation and safety assessment. AP1000 safety margins are large and well established and the chances of accidents that could put public in danger are very low. Simplicity in design has made it less expensive to build, operate, and maintain. Under accidental condition, passive safety systems depend upon natural circulation, gravity, compressed gas and evaporation to provide for long term cooling.

Normal steady state operation at full-power is typical for most of pressurized water reactor systems with respect to thermal hydraulic phenomena. However, most important feature in the designing of AP600 and AP1000 is

that they utilizes core residual heat to exploit the following core cooling circulation functions [9]:

Natural circulation in main system, passive residual heat removal loop circulation, core make-up tank loop circulation loops, lower containment sump recirculation loops, containment steam internal circulation and containment external air circulation. The special containment performing the function of containment heat removal has been represented in Figure 2-2 and the Figure 2-1 describes the connections of the primary system with passive safety systems and it consists of:

 A passive residual heat removal system

 Core make-up tanks

 Automatic de-pressurization system

 Accumulator tanks

 Containment refueling water storage tank

 Lower containment sump

 Passive containment cooling system

Figure 2-1 AP600/AP1000 containment cooling system [16]

Figure 2-2 AP600/AP1000 passive safety systems [16]

2.1.5 Hybrid passive safety systems

Switching from active safety system to a passive safety system requires extensive modifications in the plant design; rather some nuclear industries of Japan have applied the idea of hybrid passive safety systems without going directly to a complete passive system. In a research, hybrid type safety systems have been conceived by Mitsubishi for future pressurized water reactor [18]. It consists of active and passive safety systems grouping, to achieve improved safety, high reliability, and better economics. It includes application of the passive safety components using natural circulation providing force. The passive safety components and systems are taken against conventional safety systems consisting mainly of active components, as means for further safety of the nuclear power design. Mitsubishi has succeeded in introducing the passive safety concept to current active safety systems and has structured the hybrid safety systems efficiently due to weak driving forces.

The benefits of active safety systems are its effectiveness for controlling and limiting the accident. It allows the operator to select the best possible way for accidental condition termination.

The hybrid systems depend upon the passive safety systems for LOCA related accidents that have low probability of occurrence. The active safety systems are used for other accidents like blackout or breakage in secondary piping and for very small LOCA. Moreover, active safety systems act as backup of passive safety systems under accidental conditions, in case the active safety systems do not operate due to some technical problems. Such optimized combination can improve safety and reliability while maintaining the advantages in the present systems. In the present PWR design, the active safety systems consist of charging system, safety injection system and feed water system and their drive sources.

As mentioned above, most of the research work has been performed with reference to large size reactors. Some of these reactors are operating in the different parts of the world and some are planned to be commissioned in near future. The research related to implementation of the passive systems in small scale pressurized water reactor design is not completed yet. Therefore, the research we have made is an effort of implementing the aforementioned passive systems in small scale advance pressurized water reactor design. The designing of this proposed passive system is a milestone in this direction. This proposed system could be a potential candidate for incorporation in advanced small scale pressurized water reactors.

Chapter 3

3 Mathematical modeling

Safety and accident analysis with reference to nuclear power plant require better knowledge of thermal-hydraulic phenomena related to reactor anomalies. It is a two-phase phenomena and its modeling involves three types of approaches. These approaches appear in the form of three models called three-equation model, five-equation model and six-equation model. Each model employs three types of fundamental equations of mass, energy and momentum balances. The mass balance equation keeps track of thermal hydraulic system inventory. The energy balance equation performs its role in balancing the energy of thermo-hydraulic system. It has positive or negative sign depending upon whether heat is entering or leaving the system. Momentum is a vector quantity and it must be conserved during a thermo- hydraulic flow process. Momentum balance equation plays its role by balancing the momentum terms such as pressure, gravity and shear components.

The selection of thermo-hydraulic flow models is dependent on different domains of nuclear power plant. For a typical nuclear power plant the model selection depends upon the local thermo-hydraulic conditions. It is well known that during normal PWR operation, sub-cooled conditions are normally observed in main coolant loops. However, in the upper portion of reactor core nucleate boiling is allowed. In pressurizer, saturation conditions are generally maintained. In steam generators sub-cooled water circulates in tubing connected to the primary side and in the secondary side, saturation conditions are maintained for the production of steam. Therefore in normal PWR operation homogenous model is sufficient for reactor simulation. However in

accidental cases, large amount of voiding can occur at specified places and the three equation model may not be sufficient for accident analysis. In such cases five or six equation model is considered for simulation. But due to large applications domain, five-equation model is preferred over six equation model in order to see the overall behavior of the plant rather than small number of components. The scope of simulation for five equation model includes main coolant system and proposed passive system and rest of the systems has been simulated using three equation model, these systems include charging system, safety injection system; accumulator injection system and residual heat removal system. The detail of three and five equations models is stated below.

3.1 Three equation model

In this model, the different fluid phases are considered as a homogeneous mixture. The mass, momentum and energy conservation equations are solved for the mixture. In this study two phases (water and steam) of fluid are considered and it is assumed that these two components are evenly mixed and both have equal velocities and temperatures.

A detailed derivation of this model is available in literature[19], however the summary of its main equations is given below:

3.1.1 Mass balance equation

The mass balance equation keeps track of the mass conservation in the thermodynamic flow system and is given in most general form as basic equations of continuity in literature [20]. However as a special case in one- dimensional flow these equations can be written as:

∂ ∂ (3.1) ρ + (ρ v ) = 0 ∂t m ∂z m m

Where vm and ρm are the mixture velocity and density. With the assumption of homogeneous flow the density of the mixture is given in equation 3.2:

ρm = αρv + 1 − α ρl , (3.2)

ρv , ρl are vapor and liquid densities; α , the void fraction is given as:

V (3.3) α = V V

Equation 3.1 can be written as:

∂ ∂ (3.4) ρ + G = 0 ∂t m ∂z m

where Gm represents mixture mass flux.

3.1.2 Momentum balance equation

The momentum balance equation in axial direction is obtained by equating rate of change of momentum equal to the applied force:

∂ ∂ ∂p (3.5) ρ v + ρ v 2 = − + ρ g + F + f ∂t m m ∂z m m ∂z m wm m where p is the pressure of the mixture, g is acceleration due to gravity,

Fwm is mixture wall friction and fm is the change in momentum term due to presence of some components such as pump, valve or some loss coefficient.

The above equation can be written as:

2 ∂ ∂ Gm ∂p (3.6) Gm + = − + ρm g + Fwm + fm ∂t ∂z ρm ∂z

3.1.3 Energy balance equation

The energy balance equation for homogenous model is given by:

∂ ∂ ∂p (3.7) ρ h + (ρ h v ) = + q + ρ v g ∂t m m ∂z m m m ∂z wm m m

Where hm , qwm are mixture enthalpy and wall heat transfer

The equation 3.7 can be written as

∂ ∂ ∂p (3.8) ρ h + (G h ) = + q + ρ v g ∂t m m ∂z m m ∂z wm m m

Using this homogeneous model auxiliary system of reference power plant including charging system, safety injection system, accumulator injection system and conventional island of reference power plant has been simulated.

3.2 Five equation model

This model is based upon five equations, two equations for mass of the liquid and gas, two equations for energy of liquid and gas each and one equation for momentum of the mixture of two phases. This model is used in nuclear safety analysis for large simulation domain [21]. The pressure distribution, volumetric flows and enthalpies of the phases are solved using these five equations. For formulating momentum equation, two phases are considered as homogeneous which is based on assumption that the information required in many fluid engineering problems is often the response of mixture rather than individual components [22].

A drift flux model [23, 24] is used for finding the velocities of the phases involved in many thermo-hydraulic problems. The relative velocity between phases is called slip velocity and dependent upon flow variables[25]. The

most important aspect of using this model is the reduction in field equations required in comparison with pure two-fluid model. The detail of the five equation model is given below:

3.2.1 Mass balance equations

The mass of the thermo-hydraulic system is represented by solving the following one dimensional equations for liquid and vapor phases:

∂ ∂ (3.9) (α ρ ) + (α ρ v ) = m ∂t l l ∂z l l l l

∂ ∂ (3.10) (α ρ ) + (α ρ v ) = m ∂t v v ∂z v v v v

where

αv , αl are phase fractions of vapor and liquid, ρv , ρl are density of vapor and liquid, m v, m l are interfacial mass transfer of vapor and liquid, vv , vl are velocities of two phases and following relations connect the two phases,

αv = 1 − αl = α (3.11)

m v = −m l = m (3.12)

Velocities of two phases are considered to be equal i.e. vv = vl.

3.2.2 Momentum balance equation

Momentum equation for this model is the same equation of mixture momentum equation as taken in homogeneous model and is given below:

∂ ∂ ∂p (3.13) ρ v + ρ v 2 = − + ρ g + F + f ∂t m m ∂z m m ∂z m wm m

Where ρm is given in equation 3.2. The solution of momentum equation has an upper bound of critical mass flow. The mixture frozen sound speed as given below is used as a limitation for critical mixture velocity[19]. Where

ρ and α have their usual meaning as given in equation 3.2 and 3.3 and C푙 and C푣 are the velocities of the two phases under extreme conditions as defined in equations 3.24 and 3.25.

1/2 (1 − α)ρ 1 1 (3.14) Cfr = − 2 − 2 ρl Cl Cv

1 휕ρl (3.15) 2 = 퐶푙 휕푝

1 ∂ρ푣 (3.16) 2 = C푣 ∂푝

If vm is greater than Cfr than vm = Cfr .

3.2.3 Energy balance equation

Two separate equations, one for vapor phase and one for liquid phase are solved as given below:

∂ ∂ (3.17) α ρ h + α ρ h v ∂t v v v ∂z v v v v ∂p = α + m h + q + q + F v + α ρ v g v ∂z v iv iv wv il iv v v v

∂ ∂ (3.18) α ρ h + α ρ h v ∂t l l l ∂z l l l l ∂p = α + m h + q + q + F v + α ρ v g l ∂z l il il wl il il l l l

Where hv, hl are enthalpies of vapor and liquid phases and qwv , qwl are wall heat flux for vapor and liquid phases, vil, viv are interfacial vapor and liquid phases velocities and Fil , Fiv are interfacial friction between two phases and is related as,

Fil = −Fiv = Fi (3.19)

vil = viv = vi (3.20)

In the present study, this model is used for simulating nuclear island systems of reference power plant including reactor coolant system, main steam system, feed water system and proposed passive safety injection system etc.

3.3 Thermo-hydraulic correlations

The wall friction of mixture, wall heat transfer for liquid and gas phases, interfacial mass and heat transfer need constitutive equations in addition to the basic conservation equations. The discretized forms of the these conservation equations are solved via sparse matrix solution giving pressure, liquid enthalpy and vapor enthalpy distribution in nodes, the mixture velocity distribution in branches and void fraction distribution in nodes. The void fraction includes the effect of the phase separation, which is calculated by drift flux formula.

The pressure and mixture velocity distribution includes the effect of wall friction which is included in momentum equation. The Blasius correlation presented in equation 3.21 is applied for turbulent flow and the equation 3.22 is applied for laminar flow regimes[26].

−0.25 Fwm = 0.316 Re (3.21)

64 (3.22) F = wm Re

Where, Re is the well known Reynolds number.

All heat transfer correlations are selected on the basis of critical heat flux, Leiden frost temperature, wall temperature and saturation temperature. The forced convection for single phase is solved by Dittus-Boelter correlation[19].

k (3.23) h = 0.023 Re0.8Pr0.4 D

Chen correlation [27-29] is selected for boiling heat transfer phenomena. For the calculation of critical heat flux (푞푐푟 ), Westinghouse correlations [28], Biasi correlation [19] and Zuber-Griffith correlations [30] are used under their applicable ranges. The additional heat transfer for the dry wall due to the annular inverted film boiling is found using Bromley correlation [31] .

3.4 Decay heat

The evaluation of heat generation in reactor core after shutdown is necessary for determining the cooling requirements under accidental conditions. Reactor shutdown heat generation is the sum of heat produced due to fissions from delayed neutrons and decay of fission products. These two sources initially contribute equal amount to the shutdown heat generation however within

minutes, shutdown fissions from delayed neutron emission are reduced to a negligible amount.

3.4.1 Fission heat after shutdown

The heat generation from fission by delayed neutrons is obtained by solving the neutron kinetic equation after a large negative reactivity (assuming a single group of delayed neutrons). The time dependent neutron flux can be given by:

β−ρ β ρ − t (3.24) φ t = φ [ e−λ1t − e l ] 0 β − ρ β − ρ

Where φ0 is steady-state neutron flux prior to shutdown and β is total delayed neutron fraction and ρ is step reactivity change, l is prompt neutron life time, λ1 is decay constant for longest lived delayed neutron precursors and t is the time after reactor shutdown. For U235 fuel, water as moderator,

-1 -4 these parameters are: λ1 is 0.0124 s , β=0.006 and l=10 . When put into above equation for a reactivity insertion of ρ=-0.09, the fractional power which is proportional to the flux and is given by:

Q (3.25) = 0.0625e−0.0124t + 0.9375e−960t Q0

Where t is in seconds, the second term in above equation decreases sharply and becomes negligible in less than 0.01 second. Consequently, the reactor power decreases exponentially over a time period of 80 seconds, which is about the half life time of longest lived precursors.

3.4.2 Fission product decay Heat

Fission product decay heat generation after reactor shutdown is major source of heat. The empirical relation for rate of energy release due to β and γ emissions from fission products is given by:

β energy release rate = 1.40 t -1.2 MeV /fission s (3.26)

γ energy release rate = 1.26 t -1.2 MeV /fission s (3.27)

In the above expressions, t is time in second after fission. These equations are valid within a factor of 2 for 10 s

The residual heat at time τ seconds after reactor startup due to fissions occurring during the time interval between 휏 and 휏 + 푑휏 is given by:

−1.2 10 ′′′ 3 dPβ = 1.40 τ − τ 3.1 × 10 q0 dτ Mev/cm s (3.28)

−1.2 10 ′′′ 3 dPγ = 1.26 τ − τ 3.1 × 10 q0 dτ Mev/cm s (3.29)

′′′ Where q0 = volumetric heat generation rate.

For a reactor operating at a constant power level over the period, 휏푠 we integrate above equations to get decay heat from all fissions:

11 ′′′ −1.2 −0.2 3 Pβ = 2.18 × 10 q0 [ τ − τs − τ ]Mev/cm s (3.30)

11 ′′′ −1.2 −0.2 3 Pγ = 1.95 × 10 q0 [ τ − τs − τ ]Mev/cm s (3.31)

These equations can be represented as fraction of operating power level (P0).

P β −1.2 −0.2 (3.32) = 0.035[ τ − τs − τ ] P0

P γ −1.2 −0.2 (3.33) = 0.031[ τ − τs − τ ] P0

Total fission power decay heat rate is:

P −1.2 −0.2 (3.34) = 0.066[ τ − τs − τ ] P0

It is assumed that all the energy from beta particle is deposited in the fuel whereas only a fraction of gamma ray energy is added in the fuel. The rest of the gamma ray energy is deposited in the structural material of the core and surrounding supporting structure.

3.5 Advance process simulation environment

Mass, energy momentum equations appearing in previous sections can be solved with the help of different thermal hydraulics codes available for simulation purposes. Advance process simulation environment called APROS is one of such kind of code which can be used for the solution of thermo hydraulic model. This code has certain advantage over other software‟s available in the market. This is modular software and has been used extensively in nuclear power industry [32-35]. It can simulate nuclear power plant systems for electrical systems, core neutronics, thermal hydraulics and instrumentation & control. It provides a simulation environment for construction of a thermal hydraulic process model by connecting components from a given library.

APROS code has been used extensively for feasibility studies, design and safety analyses, testing of operating instructions, accident analysis,

optimization and training. For reference power plant simulation, this software has been used for its one dimensional water and steam flows for homogenous and drift flux models. In its simulation model, thermal hydraulics has been described using fundamental equations of mass, momentum and energy and its related correlations for heat transfers and friction. The thermal hydraulic solution models exploit staggered discretization scheme for its homogenous and five-equation models. In it, enthalpy and pressure are calculated at each node. The branches contribute for the calculation of flow related parameters. The water and steam properties are functions of enthalpy and pressure which are used in already stored tables. These models have the effects of valves and pumps as well as perform boron concentration calculations. A model for transport of chemicals, non-condensable gases and radioactive species is used in these two phase flow models. These models also include correlations according to the plant modeling requirements. The heat conduction modules can be connected to each thermal hydraulic model. The heat flow between fluid and solid heat structures is found using empirical correlations. The heat transfer phenomenon is defined on the basis of saturation temperature, wall temperature and fluid temperatures. Separate sections of a system can be simulated with different thermal hydraulic models. The different models can be connected using special connections. APROS uses an iterative solution method for solution of equations.

There are two different process simulation levels in APROS for research work. Development of the simulation is started with the definition of the necessary connection points. Between any two points the user adds desired process component such as pipe, tank or valve. According to the given data, the code automatically creates the necessary nodalization and calculation level modules. During simulation development, a user can choose between homogenous and drift-flux model.

The validation of APROS has been performed with experimental as well as with authenticated codes. For this purpose, PACTEL test facility and computer simulations of selected experiments were performed [36, 37].

In a validation activity, APROS-based analyzer for kola nuclear power plant [38] and Loviisa nuclear power plant analyzer have been established [39]. The model analyzer for Kola nuclear power plant include: six loops of the power plant, reactor core, steam generators, secondary circuit model, and automation system. On the basis of the tests conducted on these analyzers it can be concluded that APROS is a powerful and flexible tool for the effective exploration of wide spectrum of problems in dynamic processes modeling and in the area of safety analyses both for nuclear plants and other power installations.

In another study the detailed thermal hydraulic modeling of the PWR pressurizer using both APROS simulation software package and TRACE code has been conducted [29]. The constructed models have been assessed against data from in-surge, out-surge and spray experiments conducted at test facilities suitable for separate effect pressurizer tests in order to find an optimal modeling approach for simulating pressurizer transients. The main goal of the research was to validate the pressurizer model of APROS against the background provided by experiments from three different test facilities, namely PACTEL, MIT and NEPTUNUS facility. In addition, a model of Loviisa nuclear power plant pressurizer has been developed and the pressurizer behavior during a turbine trip transient was simulated for the purpose of investigating the scaling effect on the performance of the APROS pressurizer calculation. The separate tests have been simulated using different nodalization schemes and modeling options, and extensive comparisons between the results given by APROS and TRACE have been made. The TRACE simulations are also a part of the international CAMP code validation program of the United States nuclear regulatory commission. Based on the simulation

results, an evaluation of the capabilities of the heat transfer models to predict pressure behavior in the two codes has been provided. In this validation study, three experiments were conducted to study the behavior of the new horizontal steam generator construction of the PACTEL test facility[34].

Following are selected validations cases performed during APROS development[40]. The test are “Homogeneous model validation”, “Hungarian PMK test facility (PWR): SBLOCA”, “Edward's pipe: blowdown of horizontal pipe”, “Top blowdown test facility (Battelle Frankfurt)”, “OECD ISP 6, steam blowdown”, “Marviken critical flow test MXC-17”, “FRIGG-loop, two-phase heat transfer”,Two-phase model (five-equation model), “homogeneous and five-equation thermal hydraulic models, TVO nuclear power plant compared to accident analysis code results (GOBLIN, BISON)” etc. More information regarding these validation test is available in literature[41]. The five equation model is also been used in the development of SMABRE code developed by VTT, a technical research centre of Finland.

Chapter 4

4 Nuclear safety systems

Safety and availability are prime factors for nuclear power plant operation. Safety has more importance due to public and capital investment protection that is why safety first policy is adopted for nuclear power plant operation. The most crucial and critical component of a reactor is reactor core. Therefore, the reactor protection is actually attributed to the safety of reactor core. Nuclear safety systems are particularly designs and play vital role in core protection particularly under accidental conditions. During normal operation of a plant, highly pressurized water circulates through the reactor core for removal of generated heat. There is a probably of the coolant leakage from the reactor coolant system pressure boundary. The leakage from reactor coolant system has terrible effect upon the reactor core performance. Due to leakage, the water inventory decreases and the core cooling is affected. The higher the break size, the more rapidly it will empty starting from the pressurizer. Under such conditions, water is injected in the reactor core by emergency core cooling system whose detail is given in the next section.

4.1 Emergency core cooling system

As mentioned earlier, a number of engineered safety features have been put into the plant design to decrease the effects of LOCA. Protection of reactor core in emergency situations is the main function of this system. It utilizes some systems of the plant. For a typical pressurized water reactor design, the emergency core cooling system consists of charging system, boron injection system, safety injection system, accumulator injection system, residual heat

removal system [15]. The detail of these systems is available with the literature; however a brief summary is given below:

4.1.1 Charging system

This system is a part of chemical and volume control system and is used to maintain the required water inventory in the reactor coolant system during normal operation of plant [42]. Normally one charging pump is working for its charging necessities. The instrumentation is implemented in such a way that on demand (receipt of the safety injection signal) second charging pump is started automatically and inserts borated water into cold leg as shown in Figure 4-1. The charging pumps are only used in direct injection phase. When reactor coolant system pressure decreases, the pumps run in their overflow rate operation. Due to low discharge pressure under accidental condition, the protection system makes the charging pumps trip. A good application of this system is available with very small break LOCA with details given in section 4.1.6.4.

Figure 4-1 Block diagram of charging and boron injection system[42]

4.1.2 Boron injection system

The main function of the system is to provide necessary boron to the reactor core through cold leg of the main coolant system. This system consists of two boron transfer pumps and a boric acid tank as shown in Figure 4-1. In case of LOCA, the boric acid transfer pumps are stared automatically. The pumps draw boric acid solution from boric acid storage tank and deliver it to the suction of charging pumps. The charging pumps deliver this borated solution to reactor core through reactor coolant system cold leg. Latter on boron from refueling water storage tank is supplied through charging, safety injection system and residual heat removal system pumps. Accumulator tanks also supply borated water to the reactor core when in operation.

4.1.3 Safety injection system

The function of this safety system is to provide core cooling in the event of LOCA. Under accidental condition this system should meet the “Acceptance criteria for emergency core cooling system for light water reactors [43].The system comprise of four parallel pumps as represented in Figure 4-2. The suction lines of these pumps are connected to refueling water storage tank and discharge lines are connected to cold leg through residual heat removal system. Each discharge line has two branches: one is connected to cold leg via accumulator injection line and other is connected to hot leg.

Each pump is provided with a mini flow line for pump protection. The valve in this line closes automatically when switching from injection phase to recirculation phase. During long term cooling phase, the recirculation flow is redirected from the cold leg to hot leg and vice-versa to prevent deposition of boron in the core area.

Figure 4-2 Block diagram of safety injection system[43]

4.1.4 Accumulator injection system

This system is normally used in large break LOCA. It consists of two accumulators; one accumulator is connected to loop-A and second accumulator is connected to loop-B as shown in Figure 4-3. Each accumulator is filled with borated water and pressurized with nitrogen gas. After LOCA, as the main coolant system pressure decreases, the pressurized gas in the accumulator would expand adiabatically and force borated water from accumulator into reactor core. It is assumed that the entire water volume of one accumulator would spill to the containment through postulated break. The other accumulator is sufficient to fill up the reactor vessel lower plenum and down-comer and to initiate the re-flooding of the reactor core.

Figure 4-3 Block diagram of accumulator injection system[43]

4.1.5 Residual heat removal system

The normal operation of this system is to remove residual heat from reactor core during normal plant shutdown. In LOCA case, this system is used to remove decay heat from the reactor core during low pressure. This system consists of two pumps, two heat exchangers, associated piping and valves necessary for its working. The discharge line of each pump has three branches. One discharge line is connected to cold leg through accumulator discharge line. Second discharge line is connected to hot leg. Third discharge line is connected to the suction of safety injection system pump discharge line as shown in Figure 4-4.

Figure 4-4 Block diagram of residual heat removal system

During direct injection phase, the pumps take suction from refueling water storage tank and deliver borated water to cold legs. During recirculation phase, they take suction from containment sump and deliver the containment water to cold leg. In simultaneous recirculation phase, these pumps deliver borated sump water to hot or cold leg alternatively.

If required, during recirculation phase, the heat exchangers of this system provide services for containment spray system for reducing temperature and pressure of the containment rapidly.

Figure 4-5 Block diagram of emergency core cooling system

4.1.6 Role of ECCS during various LOCAs

The response of emergency core cooling system, represented in Figure 4-5, is different for different break sizes. The spectrum of breaks in the main coolant system has been represented in Table 2-1.

4.1.6.1 Accident description of large break

A large break LOCA covers equivalent diameter greater than 243 mm up to and including double ended rupture. In this accident, a special signal called safety injection signal is generated from pressurizer low pressure signal as shown in Table 4-1. This signal is an indication that accident has accrued on the nuclear power plant and it gives starting signals to emergency core cooling system. On receipt of this signal emergency reactor trip is initiated and emergency diesel generators are started. The detail of this accident is available in the literature [13]. A brief response of emergency core cooling system is given below:

 Charging pumps are started

 Safety injection system pumps are started

 Residual heat removal pumps are started

Emergency core cooling system response in such breaks represents three phases whose detail is given below:

In direct injection phase, charging pumps inject borated water into cold legs from boric acid solution tank and then from refueling water storage tank. Safety injection system pumps insert cooling water from refueling water storage tank when pressure in the cold leg is less than their shut-off-head. As the pressure in the main coolant loop decreases below normal gas pressure in the accumulator, its injection is started. Residual heat removal pump injection is started when pressure has decreased below their shut-off-head.

Recirculation phase is started when refueling water storage tank is emptied. In this phase, safety injection system pumps take suction from residual heat removal system pumps and residual heat removal system pumps take suction from containment sump.

After 24 hours of the accident, the operator may start alternate recirculation phase by switching the cooling water from cold leg to hot leg simultaneously in order to remove deposition of boron from the core area. The system remain in this area for long time until pressure of the reactor coolant system is decreased and residual heat removal system takes full control of the cooling services.

4.1.6.2 Accident description of intermediate breaks

An intermediate break covers an equivalent diameter lying between 25 mm to 243 mm. In such accidental case, the SI signal is generated later than for large break. Consequently the reactor trip occurs later than for large break. In

this situation, containment set-point would not reach and spray would not actuate. The cool-down by steam generator would reduce pressure and thus leakage flow rate, increases injection flow and refill the reactor coolant system. Once the core is completely re-flooded and reactor coolant system pressure is stabilized because of the equilibrium between leakage flow and injection flow. Cool-down continue and the system comes to sub-saturated condition. For smaller breaks in intermediate break-spectrum, the reactor would be cooled through steam generator than by residual heat removal system.

For larger breaks in intermediate break-spectrum, safety injection system operation would be same as for large breaks. Accumulator injection would initiate. However, for smaller breaks with an equivalent diameter of 65 mm and below the core comes into safe condition before accumulator injection[13].

4.1.6.3 Accident description of small breaks

A small break covers breaks with an equivalent diameter lying between 25 mm to 9.5 mm. In this break, the charging pumps would not be able to compensate for the coolant lost from the break. Consequently, safety signal is generated from pressurizer low pressure. This signal will make the reactor trip, if not tripped already.

For this accidental condition accumulator injection is not required. The depressurization would be achieved by shutting down all the safety injection pumps by operator action. Running of charging pumps are continued till spillage through the break would be made up by residual heat removal pumps and reactor coolant pressure would be stable.

4.1.6.4 Accident description of very small breaks

A very small break covers breaks with an equivalent diameter less than 9.5 mm. The emergency core cooling system response for this type of break would be such that the charging pumps would compensate for the expelled coolant. The pressurizer pressure would not drop too much and safety injection signal would not actuate. Therefore, safety injection system would not operate and residual heat removal system will perform normal shutdown operation.

4.2 Simulation of emergency core cooling system

Simulation of emergency core cooling system requires a simulated environment in the form of a reference power plant. For this purpose a simulated model of reference power plant has been developed by simulating main systems necessary for working of the plant and the systems which have direct relevance with the research. All other systems of the plant have been simulated and considered as simplified. Particular attention has been focused on the simulation of emergency core cooling system such as Safety Injection System, Accumulator injection system, Charging system, Residual Heat Removal System, and Reactor Coolant System. The simulation of these systems has been performed within the frame work of their respective instrumentation & control and incorporated appropriate electrical devices for provision of electricity in pumps and valves.

For proper functioning of different sophisticated components of reference power plant, control systems have been simulated and implemented as per plant requirements. These control systems include: pressurizer liquid level control system and pressurizer pressure control system etc. For safety point of view, following control safety systems have been simulated and implemented in simulated plant model: emergency trip system and engineered safety actuation system[15].

The secondary side of reference power plant has been simulated representing heat sink for nuclear power plant primary side and effectively removes generated heat in reactor core with coolant as per plant requirement. Some of the systems in the secondary side are: main feed water system, auxiliary feed water system, main steam system, turbine bypass system, feed water system, circulating water system and electrical system. The detail of these systems is available in the documentation of reference power plant[15].

During reactor operation, the reactor core represents source of heat, its detail is available with documents of reference power plant[15]. It is modeled in one dimensional case to represent source of heat during normal plant operation. After shutdown of the reactor, the normal heat source is terminated and the core model represents decay heat.

For activating emergency core cooling system and other safety related systems, a special signal called, safety injection signal is required. This safety injection signal is generated due to any of the following reasons as represented in Table 4-1.

Table 4-1 Safety injection signal generation parameters Manual - High containment pressure 0.0294 MPa Low steam pressure 3.92 MPa Low pressurizer pressure 12.5 MPa High steam Flow with Low-Low Tavg 120%, 264 °C

4.2.1 Simulation of reactor trip system

Engineered safety features provides automatic protection against unsafe and improper operation during steady state and transient operation and generates initiating signals to mitigate the consequences of faulted conditions[15]. It ensure that the reactor safety limits are not exceeded and to accommodate

them before leading to more severe conditions. This system comprises of reactor trip system and engineered safety feature actuation system. The reactor trip system automatically prevents operation of the reactor when an unsafe condition is reached. The reactor trip system keeps watch on variables which are directly related to equipment mechanical limitations and also on variables which affect the heat transfer ability of reactor core. The reactor is made to shutdown in order to protect damage to fuel cladding or loss of system integrity which could go in the direction to release radioactivity into environment water or air. A list of reactor trips is provided in Table 4-2.

Table 4-2 A list of reference power plant trips[15] No Trip description Set point 1 (a) High neutron flux (power range low set-point) 25% full power (b) High neutron flux (power range high set-point) 109% full power 2 Intermediate range high neutron flux 25% full power 3 Source range high neutron flux - 4 Power range high positive neutron flux rate 5%/sec full power 5 Power range high negative neutron flux rate -5%/sec full power 6 Over temperature ΔT - 7 Over power ΔT - 8 Pressurizer low pressure 13.1 MPa 9 Pressurizer high pressure 16.4 MPa 10 Pressurizer high liquid level 8.12 m 11 Low reactor coolant flow 90% rated flow 12 Reactor coolant pump breakers open - 13 Reactor coolant pump bus under voltage 77% rated voltage 14 Reactor coolant pump bus under frequency 47 Hz 15 Reactor coolant pump bus under speed 89% rated speed 16 Low-low steam generator water level 8.892 m 17 Steam /feed-water flow mismatch in coincidence 9.45 m with low steam generator water level 35% 18 Safety injection - 19 Turbine-generator trip - 20 Manual -

A reactor must be safe in all of its six operating modes as given in Table 4-3. When certain parameter exceeds its limit set-point, the reactor trip system is activated and reactor is tripped. But in case of an accident, this system and engineered safety features are activated. These safety features are used for mitigating the consequences of the accident in case of design basis accidents. It automatically initiate the reactor trip if not tripped already so as to avoid accident spreading, core burning out and radioactive materials releasing to the environment. The primary requirement of engineered safety features actuation system receive input signals from the various on-going processes within the plant and containment, and automatically provide timely and effective output signals to actuate various components and subsystems. The engineered safety feature actuation system uses selected plant parameters, determines whether or not predetermined safety limits are being exceeded and, if they are, the system sends actuation signals to those engineered safety feature components whose aggregate function best serves the requirements of the accident.

The plant parameters monitored by safety feature components system are: pressurizer pressure, steam-line pressure and flow, steam generator water level and feed water Flow, reactor coolant Tavg, containment pressure, Loss of offsite power, gamma exposure rate high-high in containment air and activity high-high in containment air. These parameters have been chosen for the significant variation during severe kind of accidents. During the reactor operating conditions, if safety monitoring variables reach or exceed the preselected set point, the engineered safety feature actuation system will automatically initiate the following safeguard actuation. Safety injection, containment spray, steam-line isolation, feed water isolation and containment isolation auxiliary feed water pumps startup. Engineered safety features actuation system also has the provision for manually initiating these functions. All of the above generated signals have been implemented in the activation of

different system, necessary for working to protect reactor core under accidental condition normally called design basis accidents.

Table 4-3 Operating modes of reference power plant [15] Mode Reactivity Thermal Coolant average Coolant (keff) power temperature (°C) pressure Power operation 1.0 >2% 280-302 15.2 MPa Hot zero power 1.0 ≤2% ≥ 280 15.2MPa Hot shutdown ≤0.98 0 ≥ 280 15.2MPa

Intermediate ≤0.98 0 280 > Tavg≥ 180 15-2.94 MPa shutdown stage A Intermediate ≤0.98 0 180 > Tavg≥ 93 ≤2.94 MPa shutdown stage B Cold shutdown ≤0.98 0 ≤93 0.098 MPa Refueling < 0.95 0 ≤60 0.098 MPa

4.2.1.1 Simulation of charging system

Simulation of charging system is performed in such a way that it represents its role during normal plant operation as well as in accidental operations as per plant requirement. The simulated diagram is available in Figure A-2.

4.2.1.2 Simulation of boron injection system

This system has been simulated to provide necessary Boron to the reactor core under steady state as well as under accidental conditions. In the simulated model this system provides necessary boron to the reactor core through main coolant system cold leg. This system comprises of two boron transfer pumps and a boric acid tank. In the event of a LOCA, the boric acid transfer pumps are stared up automatically on receipt of a safety injection signal, and draw boric acid solution from boric acid storage tank and deliver it

to the suction of charging pumps. These charging pumps deliver this borated solution to reactor core through reactor coolant system cold leg.

4.2.1.3 Simulation of safety injection system

The detailed functionality of safety injection system with its process and instrumentation diagrams can be viewed from the reference power plant manual [13]. The simulated Safety Injection System performs its function of decay heat removal from reactor core during loss of coolant accident and represents direct injection phase, recirculation phase and simultaneous recirculation phase as per plant requirement. A list of simulated equipment for SIS system is listed in Table A-2 which includes: accumulators, refueling water storage tank and pumps etc. All the necessary piping and valves essential to cover the scope and extent of simulation have been simulated and its simplified simulated flow diagram is given in Figure 4-6.

Figure 4-6 Safety injection system simulation diagram

4.2.1.4 Accumulator injection system simulation

Accumulator injection system has been simulated to inject borated water into reactor core under LOCA conditions when the pressure has dropped below 5 MPa. Nitrogen gas cylinders are used to represent as cover gas pressure in the accumulators. This system is mainly used in large break LOCAs to refill the pressure vessel [41]. The simulated system performs its function as per plant requirement and consists of two accumulators, nitrogen system, piping and valves necessary for its operation. The simplified flow diagram is given in Figure 4-7.

4.2.1.5 Simulation of residual heat removal system

This system is modeled to represent the following features: removal of residual heat from the reactor core, delivering of the refueling water, circulating and purifying the reactor coolant during startup and shutdown, to be used, at LOCA, as a Low Head SIS and to perform cooling of containment sump water, during the long-term containment spray actuation [44]. Its actual flow diagram is given in reference document [15] and corresponding simplified simulated process flow diagram is given in Figure 4-8.

Figure 4-7 Accumulator injection system simulation diagram

Figure 4-8 Residual heat removal system simulation diagram

During normal plant operation, the residual heat removal system is not in operation but aligned as low head safety injection sub-system. During LOCA it performs reactor core cooling by delivering borated water from refueling water storage tank or from containment sump. The scope of simulation can represent direct injection, recirculation and simultaneous recirculation phases. It assists the containment spray system by cooling containment recirculation sump water during the long term containment spray recirculation mode following LOCA by utilizing residual heat exchangers. A list of simulated equipment for the residual heat removal system is listed in Table A-7 and the status of residual heat removal system during various operational modes as per plant requirements is represented in Table A-8 through Table A- 11.

The following variables are very important in order to see the overall behavior of the plant. Moreover, these variables are also monitored by different control systems; therefore their existence in the simulation is necessary for steady state and accidental conditions of the plant. For this purpose, the scope of simulation includes main variables used to represent associated equipment in primary and secondary systems are simulated and these variables are presented in Table 4-4.

Table 4-4 List of important variables represented in simulation # System Component Variable 1 Reactor core Core model thermal power of reactor 2 Reactor core Core model reactor vessel water level 3 Main coolant system pressurizer pressurizer level, model temperature & pressure 4 Main coolant system Relief valve relief valve flow 5 Main coolant system Point model hot leg temperature for loop A and B 6 Main coolant system point model cold leg temperature for loop A and B 7 Control Tavg primary coolant average temperature 8 Main coolant system Pipe model primary coolant flow rate for loop A and B 9 Main coolant system Point model boron concentration

10 Main feed water system Point model steam pressure for each steam generator 11 Main feed water system Steam each steam generator generator water level model 12 Main feed water system Pipe model steam flow rate and temperature 13 Main feed water system Pipe model feed-water flow rate and temperature 14 Secondary system Control valve turbine governor valve model position 15 Main coolant system Pressurizer total electric power model 16 Secondary system Electrical generator frequency generator 17 Service water system Pipe model circulating water flow 18 Main coolant system Pressurizer heater power model 19 Chemical and volume Pipe model charging flow rate control system 20 Chemical and volume Pipe model Letdown flow control system 21 Emergency core cooling Pipe model Safety injection flow system 22 Residual heat removal Pipe model RHR flow system 23 Safety injection system Pipe model Accumulator injection flow- rate

Chapter 5

5 Proposed passive safety system

The proposed passive safety system is closest to the concept of pre- pressurized tank as discussed in section 2.1.3. This system consists of two parts for the two cold legs of reactor coolant system as shown in Figure 5-1. The two parts necessary for coolant injection has been shown separately in Figure 5-2 and Figure 5-3. The main function of this system is to augment high pressure safety injection system pumps. The thermal hydraulic design basis of the proposed system have therefore been selected to be similar to the high head safety injection system so that this system can provide coolant with the same pressure and appropriate flow rate in order to ensure critical parameters under safe limits. Following is the description and specifications of the proposed passive safety injection system.

5.1 Thermal hydraulic design of the proposed system

As mentioned above the proposed system has two independent parts connected to the injection lines leading to reactor coolant system loop-A and loop-B. Each train consists of a nitrogen gas pressurized closed water storage tanks: CMT-A and CMT-B, an isolation valve and two check valves with proper piping necessary for coolant delivery to the cold legs under accidental conditions. The location, elevation, configuration and volume inventory of CMTs of the proposed system is the similar as existing accumulators of reference power plant. Nitrogen supply system (nitrogen gas cylinders) has been proposed to keep these tanks pressurized during normal plant operation. Nitrogen supply system is used for maintain pressure at 10.5 MPa, which is equivalent to the head of safety injection system pump. Flow area of

each discharge line has been adjusted by running various computer simulations to obtain the optimum flow rate. During the simulations the hardware such as piping, tanks and valves etc has been chosen to be the same as in the simulation of proven reactor technology. Design parameters of the proposed system are given in Table 5-1 to Table 5-3.

During normal power plant operation, the proposed passive safety injection system is not used but remains aligned in standby mode. In this mode it is ready to discharge borated water to reactor coolant system loop-A and loop-B when the pressure in these loops is lower than the normal operating nitrogen gas pressure in each tank. In this mode, each CMT pressurized with nitrogen gas supply system and both flow isolation valves IV1-A and IV1-B are opened. To align this proposed system with the main cooling loops standard check valves have been used in the simulation. These check valves allow one directional flow. The valves are activated when the pressure at their downstream is less than the pressure at upstream. In reverse direction these valves do not operate. The four check valves, two for loop-A (CV-1A, CV-1B) and two or loop-B (CV-2A & CV-2B) perform their function of check valves as per plant requirement.

Under accident condition, the pressure of reactor coolant system decreases due to leakage in the coolant system. When system pressure reaches below nitrogen normal operating pressure, it injects borated water into reactor core through cold legs. The delivery of borated water will continue until the tanks are emptied and discharge isolation valves are closed. It has been observed that the injection flow rate decreases with the passage of time, while for existing active safety injection system, the injection flow rate increases with time. This feature of the proposed system is more realistic with reference to core decay heat removal, which decreases with the passage of time.

Nitrogen gas CHP-A

NSV1-A LC Process Variable CMT-A

LT Set-Point

Controller Output CV4-A ~ ~ 4.2 m

3.9 m IV1-A SV1-A CV2-A CV1-A

~ CV3-B Nitrogen gas

NSV1-B Process Variable LC CMT-B

LT Set-Point

Controller Output ~ ~

CHP-B

4.2 m 3.9 m SV1-B CV2-B CV1-B IV1-B

Figure 5-1 Proposed Passive Safety Injection System

CMT-A

4.2 m

3.9 m CV2-A CV1-A IV1-A

Figure 5-2 Proposed Passive Injection System Loop-A

CMT-B

4.2 m 3.9 m CV2-B CV1-B IV1-B

Figure 5-3 Proposed Passive Injection System Loop-B

AUX. SPRAY AUX.

~ ~

SPRAY LINE A

SECONDARY

SG A LOOP SG B PRESSURIZER 002A 002B

SPRAY LINE B LINE SPRAY

LOOP A LOOP B

INJECTION A INJECTION

INJECTION B INJECTION

CHARGING

INJECTION INJECTION INJECTION

PROPOSED

INJECTION REACTOR PROPOSED

PUMP B PUMP

PUMP A PUMP

SRH A SRH A ACC B ACC

SRH B SRH

MCPA MCPB

Figure 5-4 Proposed system delivery in reactor coolant system

Table 5-1 Design parameters of proposed system tanks Parameter value Pressure 10.5 MPa Height of tank 10.25 m Bottom elevation 4.2 m Liquid temperature 25 °C Liquid level 5.29 m Boron concentration 2000 ppm Volume 56.6 m3

Table 5-2 Design parameters of proposed system isolation valves Parameter value Area 6.5E-04 m2 Length 0.23 m Shut time 50 second Elevation 4.2 m

Table 5-3 Design parameters of proposed system check valves Parameter value Area 6.5E-04 m2 Length 0.2 m Elevation 4.2 m

5.2 Controls implemented

Instrumentation and control has been implemented for normal working of the proposed passive safety system under accidental conditions. For this purpose two level controllers for two different loops have been proposed as shown in Figure 5-5 and Figure 5-6. Each level control is adjusted in such a way that each of the discharge line isolation valves (IV1A and IV1B) are closed when the water in CMT tanks reached a minimum level.

Figure 5-5 Control implementation for loop-A

Figure 5-6 Control implementation for loop-B

Chapter 6

6 Proposed system applications

6.1 Validation of reference power plant model

It is necessary to check the existing safety system‟s response before checking proposed system. For this purpose, simulation with already existing safety systems was performed including nuclear and conventional island. Initially, the simulated model of reference power plant was running at its 100% steady state power for extended period of time and a snapshot was taken which is called initial condition. After that a large size break introduced in the loop “B” cold leg of reactor coolant system by opening the valve exposed to containment as shown in Figure 6-1 and Figure 6-2.

The simulation was made to run for 10 seconds, before opening the LOCA valve. Flow area of the LOCA valve has been adjusted to represent a large break. Consequently simulation was made to operate for 790 seconds and the response of different safety related parameters has been noticed. In this transient, different safety systems interact dynamically as reactor coolant system depressurizes leading to cold shut down conditions. The results are in good agreement with final safety analysis report and are summarized in our paper [45].

6.2 Proposed passive safety system response

The response of proposed safety injection system has been checked with the help of an intermediate break. This selection is based on the assumption that for small break LOCA, charging system is mainly used to inject borated water in the coolant system due to higher pressure. While in large breaks, the pressure of the reactor coolant system decreases quickly and proposed

system is bypass while other auxiliary systems including residual heat removal system are started automatically. Therefore, intermediate break size LOCA is critical study to judge the response of the proposed system. In this transient simulation, proposed passive system, passive safety injection system is employed while active safety injection system is kept as backup. However, active safety injection system can be operated manually by operator actions when required.

6.2.1 Intermediate break LOCA

The simulation was carried out at 100% steady state power level for 10 seconds. Various parameters attained their steady state values and are mentioned in Table 6-1. At t = 10 seconds, a break was triggered into reactor coolant system loop „„B” by opening LOCA-valve as indicated in Figure 6-1. The corresponding nodalization diagram of the intact systems is represented in Figure 6-2. The simulation was made to continue for next 790 seconds and sequence of events is given in Table 6-2. In the following sections, response of various parameters is discussed and represented.

Table 6-1 Steady state safety parameters values for reference power plant Safety variable Steady state value Reactor power 100% Pressurizer pressure 15.1 MPa Pressurizer level 5.4 m Loop-A hot leg pressure 15.1 MPa Loop-B hot leg pressure 15.1 MPa Steam generator-B pressure 5.8 MPa Steam generator-B level 10.6 m Accumulator-B pressure 4.5 MPa

Accumulator-B level 5.3 m CMT-A and CMT-B water level 5.3 m CMT-A and CMT-B pressure 10.5 MPa Refueling waters tank level 9.7 m Reactor core water level 2.9 m

AUX. SPRAY

~ ~

SPRAY LINE A

SECONDARY

SG A LOOP SG B PRESSURIZER 002A 002B

SPRAY LINE B

LOOP A LOOP B

INJECTION A

INJECTION B

CHARGING

INJECTION INJECTION INJECTION

PROPOSED

INJECTION REACTOR PROPOSED

PUMP B

PUMP A

SRH A ACC A ACC B

SRH B

MCP A MCP MCP B MCP

LOCA VALVE

Figure 6-1 Reactor coolant system showing LOCA-valve

Figure 6-2 Proposed system nodalization diagram

Table 6-2 Chronological sequence of events Event detail Time (second)

Start of simulation 0.0 Transient started 10.0 Occurrence of maximum break flow 12.9 Passive safety injection system initiation 14.4 (injection phase) Reactor tripped 10.6 Both reactor coolant pumps tripped 18.6 Pressurizer empty 24.8 Start of accumulator injection 94.1 Core level (minimum) 162.2 Termination of accumulator flow 155 Start of low head safety injection (RHR system) 155 Charging pumps Tripped 450 Proposed passive safety system flow termination 684.2 Transient end 800

6.2.1.1 Break flow

Reactor coolant system experienced a maximum pressure drop across break at the initial stage of the transient. As a result, the flow through the break reaches at its maximum value of 2644.1 kg/s abruptly and afterward it decreases as coolant system depressurizes as shown in Figure 6-3.

Figure 6-3 Flow through break with passage of time

6.2.1.2 Pressurizer liquid level

After onset of LOCA the volume inventory of reactor coolant system decreases. Consequently, the pressurizer liquid level decreases abruptly from its steady value of 5.4 m to zero at 27.5 seconds, as shown in Figure 6-4.

Figure 6-4 Variation of pressurizer liquid level with time

6.2.1.3 Passive safety injection system response

The injection of passive safety injection system is started after 14.3 seconds when the pressure of reactor coolant system reaches to 10.49 MPa. The passive safety injection system flow-rate reaches at maximum value in 154.5 seconds. During flow of accumulators, the passive safety injection system injection flow remains between 40 and 46.9 kg/s, consistently. At reactor coolant system pressure of 0.4 MPa, the passive safety injection system is isolated and its valves are closed. As a result, the passive safety injection system flow-rate decreased to zero at 692.9 seconds, this is shown in Figure 6-5. Under injection phase, the CMT water level drops from 5.3 meter to 0.8 meter as given in Figure 6-6.

rate (kg/sec) rate

PSIS flow PSIS

rate (kg/sec) rate

ACC flow ACC

rate (kg/sec) rate

- RHR flow RHR

Figure 6-5 PSIS, accumulator and RHR flows with time

Figure 6-6 CMT-B liquid level with time

6.2.1.4 Accumulator’s response

After the onset of transient, accumulator injection was started at 93.94 seconds. At this stage, the pressure of the reactor coolant system was 4.5 MPa (nominal accumulators pressure was 4.5 MPa). Accumulator flow-rate reached at maximum value of 550.25 kg/s at 143.9 seconds. Accumulator flow terminated at 155 seconds as the level in accumulator reaches to 0.8 m (Figure 6-5 and Figure 6-7).

Figure 6-7 Variation in accumulator liquid level with time

6.2.1.5 Residual heat removal system’s response

After the hypothetical break the pressure in the reactor coolant system has a decreasing trend. When it is depressurized to low pressure conditions, the residual heat removal system pumps started their function. The pumps started suction from refueling water storage tank and delivered borated water to cold legs. Initially, the all the pumps were delivering water in their mini flow lines and were not able to inject into coolant system due to high pressure. Residual heat removal system started delivery after 155 seconds, after the accumulator flow decreased to zero and the reactor coolant system pressure dropped to less than 1 MPa (shut-off-head of residual heat removal pumps was 0.98 MPa), as shown in Figure 6-5.

6.2.1.6 Response of reactor power

The reactor was operating at its full thermal power during steady state simulation for first 10 seconds. After initiation of the transient, the reactor tripped at 10.6 seconds, on low pressurizer pressure. After reactor trip the reactor core represents decay heat as per plant requirement, Figure 6-8.

Figure 6-8 Normalized reactor power

6.2.1.7 Response of charging system

For normal charging requirements during normal power plant operation only one charging pump was running. After initiation of the transient when the pressurizer level was not maintained, second standby charging pump was started at 13.3 seconds. Both charging pumps were tripped at 450 seconds

manually at reactor coolant system pressure of 0.52 MPa. These charging pumps flow-rate is shown in Figure 6-9. After 450 seconds, the shown flow through charging pumps is due to boron supply pumps.

Figure 6-9 Charging system response

6.2.1.8 Loop average temperature

The average of the two loops temperature is represented by the symbol Tavg. The steady state value of loop average temperature is 302 °C. After initiation of the transient it decreases due to hot coolant lost through the break. After

100 seconds the sharp decrease in Tavg is due to supply of large amount of cold water from accumulators. Its value remained fixed around 100 °C during steady operation of residual heat removal system. Steady state value of Tavg is 95 °C at the termination of the transient as shown in Figure 6-10.

Figure 6-10 Loop average and fuel average temperature

6.2.1.9 Core enthalpy and temperature distribution

Maximum radial averaged fuel enthalpy as a function of time has been represented in Figure 6-11. The corresponding fuel centerline temperature and maximum clad temperature are shown in Figure 6-12. In these graphs, the steady state value of fuel average temperature is 519 °C. It decreased abruptly after initiation of the transient at 10 seconds. A sudden decrease in fuel temperature (Tfuel) from 100 to 150 seconds is due to large accumulator delivery in the reactor core through reactor coolant system. After the execution of accumulator flow, the residual heat removal system started injection in the reactor core and Tfuel attains a value of 150 °C. Fuel has a steady value of 145 °C at the termination of the transient as shown in Figure 6-12.

Figure 6-11 Variation in fuel enthalpy with time

Figure 6-12 Maximum fuel and clad temperatures

6.2.1.10 Total makeup flow rate

After hypothetical break, accidental condition is declared in the form of a signal called safety injection signal and injection phase is started. This signal is generated on low pressurizer pressure and injection phase is started automatically. Different safety systems inject borated water in the cold legs; these systems are shown in Figure 6-13. The total makeup flow rate is sum of the flows from charging system, proposed passive system, decay heat removal system and accumulator system. The starting and stopping of flow

for different sub-systems of emergency core cooling system also represented in Figure 6-13.

Figure 6-13 Total make-up flow rate

6.2.2 Proposed system’s response for large break

The proposed passive system‟s response is different for different break sizes. For large size break, the depressurization is fast as compared to small and intermediate breaks. The flow area of the LOCA valve is adjusted in such a way that it represents large size break when opened. In this transient a large sized break is triggered in reactor coolant system loop “B” by opening the

modeled valve coupled with the containment as boundary condition as represented in Figure 6-1 and Figure 6-2. As a result the coolant discharges through the break and primary system depressurizes abruptly thereupon. When primary coolant system pressure decreases below proposed system‟s pressure, it starts coolant delivery in primary coolant system cold legs. Charging system, accumulator injection system and decay heat removal system also inject coolant in primary reactor coolant system after cold leg pressure decreases as is the case with the normal plant requirement. Hot full power of the reactor is taken as initial conditions and is summarized in Table 6-3. The simulated transient represents the following conditions: (1) Automatic initiation of the injection phase (2) Termination of the accumulator discharge flow-rate at water level of 0.8 m (3) Termination of CMT‟s discharge flows at water level of 0.8 meter. The transient was repeated with existing safety injection system as well under same conditions mentioned above and in both the cases; the transient was made to run for 700 seconds. The summary of sequence of events is given in Table 6-4.

Table 6-3 Reference power plant parameters Type of plant Pressurized water reactor Gross electric power 325 MWe Number of main loops 2 Pressurizer 1

Moderator and coolant H2O

Fuel UO2 Primary system flow-rate 24000 T/h Primary system pressure 15.2 MPa Main coolant pumps 2 Steam Generators 2 Reactor core inlet temperature 288.5 °C Average temperature in reactor core 302.5 °C Average temperature rise in core 27.9 °C

Secondary steam flow rate 1010×2 T/h Feed water temperature 220 °C Steam system pressure 5.54 MPa

Table 6-4 Chronological sequence of events Event triggered Time (seconds) Steady simulation 0.0 Transient triggered 10.0 Reactor tripped 10.8 Reactor coolant pumps tripped 10.8 Maximum break flow 11.8 Pressurizer empty 25 Initiation of injection (SIS and PSIS) 26 Accumulators started injection 44 Accumulators terminated 115 Initiation of RHR delivery 116 PSIS terminated injection 600 Transient Terminated 700

Table 6-5 Passive safety system and safety injection system comparison Safety parameters SIS PSIS Flow-rate 80.0 kg/sec 50-45 kg/sec Injection fluid pressure 10.49 MPa 10.28 MPa Injection fluid temperature 25.0 °C 25.0 °C

6.2.2.1 Larger size break LOCA results

After initiation of the transient, the reactor is tripped at 10.8 sec, low pressurizer pressure trip and latter on the reactor power has a decreasing trend representing core decay heat as represented in Figure 6-14. The flow through break is a function of flow area; opening time and pressure drop across LOCA valve. Initially there was high flow due to large pressure difference across LOCA-valve, later on, break flow decreases as reactor coolant system depressurizes (Figure 6-15). Pressurizer behaves realistically and its level drops quickly as the volume inventory of reactor coolant system decreases (Figure 6-16). The second stand-by charging pump is started (first was already in operation) for makeup purposes as per plant requirement [42] as shown in Figure 6-17. The pressure continues to decline and when it decreases below 10.5 MPa, the injection from Passive Safety Injection System (PSIS) is started automatically (pressure in each CMT is 10.5 MPa). Similar behavior is also observed with safety injection system simulation whose pumps inject cold borated water in cold legs when pressure of reactor coolant system is less than 10.5 MPa. Maximum flow through these systems is attained when the accumulators are emptied and injection lines are available for passive safety injection system or safety injection system only. Injection flow with passive safety injection system gradually decreased from 50 kg/sec and finally attained a value of 38 kg/s. It is to be noted here that passive safety injection system has higher flow rate initially and latter on its flow decreases as temperature of the core decreases. At CMT-B water low-level of 0.8 m, the passive safety injection system isolation valves were closed and the passive safety injection system flow gradually decreased to zero at 600 seconds as illustrated in Figure 6-18. Whereas, four pumps of safety injection system were kept on running until the end of transient. The CMT-B water level is dropped from 5.4 m to 0.8 m at the termination of the transient as shown in Figure 6-19. Accumulator injection is started below 4.5 MPa (4.5

MPa was the nominal pressure in accumulators). Accumulator flow was terminated at 0.8 m as is shown in Figure 6-20.

During transient simulation with passive safety injection system and safety injection system separately, Residual Heat Removal (RHR) performed its duty as per plant requirement [46] and its pumps were running in their respective mini-flow lines but were not in position to inject water in primary coolant system due to high pressure. Residual heat removal system pumps started injection after 115 seconds, when the accumulator flow decreased to zero and the reactor coolant system pressure dropped below 1MPa (shut off head of residual heat removal pump is 1MPa), as shown in Figure 6-18. During this simulation exercises in both the cases, “loop average temperature (Tavg)” and

“fuel average temperature (Tfuel)” were monitored. Tavg is the average of cold leg and hot leg temperatures per loop where as Tfuel is the average of the core temperature. Combined plot of these two parameters has been shown in

Figure 6-21. After reactor trip, Tavg decreases from 302 °C steadily and after

45.8 sec the sharp decrease in Tavg is due to accumulator cold-water flow.

After the end of accumulator flow, a little increase in Tavg has been noticed. It is more obvious with passive safety injection system simulation due to its low flow. Residual heat removal system performed well in both the cases and reduced Tavg from 150 °C to 98 °C (designed for 180 °C to 93 °C). On the other hand, the value of Tfuel decreases from steady state value of 519 °C and after 45.8 sec the sharp decrease in Tfuel is due to accumulator cold-water flow. After the termination of accumulator flow, there is little increase in Tfuel with passive safety injection system; this is due to small flow of passive safety injection system as compared to safety injection system. After the end of accumulator flow, the residual heat removal system started delivery and the fuel attain its steady value around 140 °C. At the termination of the transient, the fuel gets a steady value of 133 °C as represented in Figure 6-21.

Figure 6-14 Normalized reactor power

Figure 6-15 Flow through simulated break

Figure 6-16 Pressurizer liquid level

Figure 6-17 Charging pumps A, B flow-rates

100

Figure 6-18 PSIS, accumulator, RHR flows and cold-leg pressure

101

Figure 6-19 Response of CMT-B liquid level with time

Figure 6-20 Response of accumulator liquid level with time

102

Figure 6-21 Variation in loop average and maximum fuel temperature

6.2.2.2 Simulation for different break sizes

A comprehensive testing of passive safety injection system is performed at different break sizes. In this testing the active safety injection system and passive safety injection system are further analyzed and compared at 10%, 15%, 20%, 25% and 30% breaks. In this testing both systems performed approximately the same and the results are presented below in graphical forms:

103

100.0 100.0 10% break 15% break 87.5 87.5 20% break 25% break 30% break 75.0 75.0

62.5 62.5 Existing Proposed

50.0 50.0

SIS flow SIS (kg/sec) CMTflow (kg/sec)

37.5 37.5

25.0 25.0 0 150 300 450 600 750 0 150 300 450 600 750 Time(sec) Time(sec)

Figure 6-22 Accumulator flows comparison for a spectrum of break sizes

600 600 10% break 15% break

500 20% break 500 25% break 30% break

400 400 Existing Proposed

300 300

200 200 ACC flow (kg/sec) ACC

100 100

0 0 0 50 100 150 200 250 300 0 50 100 150 200 250 300 Time(sec) Time(sec)

Figure 6-23 Accumulator flow comparison for different break sizes

104

170 10% break 170 15% break 160 20% break 160 25% break 150 30% break 150

140 140

130 130

120 120

110 110 RHRflow (kg/sec) Existing 100 100 Proposed

90 90

80 80 0 200 400 600 800 0 200 400 600 800 Time(sec) Time(sec)

Figure 6-24 Residual heat removal system flow for spectrum of breaks

1400 1400 10% break 15% break 1200 1200 20% break 25% break 30% break 1000 1000

800 Existing 800 Proposed

600 600

400 400 Cumulative flow (kg/sec) Cumulative

200 200

0 0 0 200 400 600 800 0 200 400 600 800 Time(sec) Time(sec)

Figure 6-25 Comparison of cumulative flows for different break sizes

105

400 400 10% break 15% break 350 350 20% break 25% break 30% break 300 300

250 Existing 250 Proposed

200 200

150 150 Maximum clad temperature (°C) temperature clad Maximum

100 100 0 200 400 600 800 0 200 400 600 800 Time(sec) Time(sec)

Figure 6-26 Maximum cladding temperature for different break sizes

106

Chapter 7

7 Conclusion and future recommendations

In this research, a passive safety system has been designed and modeled for a small scale reference power plant. During simulation test run, it is observed that the proposed system performed well by protecting the reactor core, in place of an existing active safety injection system for mitigating the consequences of loss of coolant accident. To accomplish this task, a point wise description of research objectives achieved is given below:

 Modeled a reference power plant with and without the proposed passive safety injection system including all major interacting systems, e.g.:

 Nuclear Island

 Reactor coolant system

 Charging system

 Safety injection system

 Accumulator injection system.

 Residual heat removal system

 Component cooling water system

 Conventional Island

 Feed water system

 Auxiliary water system

 Main steam system

 Turbine system

107

 Proposed and simulated a passive safety injection system.

 Simulated a hypothetical accidental condition and observed the response of the active safety injection system and the proposed passive safety injection system for a range of break sizes.

 Compared the two approaches for safety augmentation.

In this study, the reactor core safety is ensured by the use of a proposed passive safety injection system as shown in Figure 5-1 in place of an existing active safety injection system by using the simulated model of reference power plant. A hypothetical accidental condition has been created for observing the response of the proposed passive system and the existing safety injection system. The results obtained were compared in both the cases from an intermediate to large size breaks in reactor coolant system cold leg.

The functionality and performance of the proposed system has been analyzed and it is found that the proposed system can handle the investigated break safely by keeping all operating parameters within limits. Under all these accidental conditions, the temperatures of the core remained within safe limits as given in Figure 6-10, Figure 6-12, Figure 6-21 and Figure 6-26.

It is to be noted that the proposed system is fast in action since no delays are involved as given in Figure 6-18. The flow-rate of the proposed system is decreasing with the passage of time and is more realistic as the decay heat is also decreasing with time as given in Figure 6-5 and Figure 6-8 respectively. At one stage in comparison (Figure 6-26), the maximum clad temperature in the core started increasing after 100 seconds due to low flow of proposed system but it is not out of range and finally normal temperature is achieved after 300 seconds.

108

The proposed system consists of passive components which are lesser in numbers as compared to existing, active safety injection system, which is a huge system and consists of many active components. There is no provision of power source for its operation and no operator actions are involved. Therefore, the proposed system is a better choice for handling DBA accident due to fewer components that resulted in simplification of the plant with improved safety.

Future recommendations

In order to enhance the research work further for a complete passive power plant following recommendations are suggested:

 With the simulated model of reference power plant and proposed safety injection system, a number of sensitive studies can be performed. Like finding volume inventories of CMT‟s for a particular break size in any of its reactor main coolant loops hot leg or cold leg.

 It is proposed to enhance the research further by incorporation of CMTs which are pressurized from reactor coolant system loops itself through its pressure balancing lines. But in our proposed system these CMTs are pressurized with nitrogen gas in order to avoid leakage due to pressure balancing lines. To check the overall behavior, simulation with pressure balancing line to CMTs may be performed.

 It is proposed to incorporate and simulate the provision of direct vessel injection, automatic depressurization system, Passive natural recirculation loops, under gravity flow to core under atmospheric conditions and sump water natural re-circulation loops. These modifications will reduce the number of components from existing plant further and will improve economics and safety for next generation simplified passive power plant design.

109

References

1. Cheng X., et al., Role of Passive Safety Systems in Chinese Nuclear Power Development. Science and Technology of Nuclear Installations, 2009.

2. Lamarsh, J.R., Introducion to Nuclear Engineering (Chapter-4). 1983: Addison-wisley publishing company.

3. El-wakil, M.M., Nuclear Heat Transport. 1971, NewYork: International Textbook Company, USA.

4. Collins, E.D., The Three MIile Island Accident and Post-accident recovery — What did leanr? 1982.

5. Milhaud, G., The lesson of the Chernobyl disaster. Biomedicine & Pharmacotherapy, 1991. 45(6): p. 219-220.

6. IAEA, Safety Related terms for Advanced nuclear reactors 1991.

7. IAEA, Natural circulation in water cooled nuclear power plants. November, 2005, IAEA: VIENA.

8. Nayak A. K. and Sinha R. K., Role of passive systems in advanced reactors. Progress in Nuclear Energy, 2007. 49(6): p. 486-498.

9. Tower, S.N., Passive and simplified system features for the advanced Westinghouse 600 MWe PWR. Nuclear Engineering and Design, 1988. 109: p. 147.

10. Dennis A., Gyanfransco S., and Mariano., Europen passive plant program A design for the 21st century. Nuclear Engineering and Design, 1998. 179: p. 17- 29.

11. Clayton J. C., The Shipping port pressurized water reactor. 1993, Westinghouse Electric Corporation.

12. Status report 75 - Advanced Passive pressurized water reactor (AP-600). 2010, IAEA: Austeria.

13. Hwang, Safety Injection system(manual). 1995, SNERDI.

14. Jones, O.C., Nuclear Reactor Safety Heat Transfer (Chapter 9). 1981: McGraw Hill

15. Zhou, FSAR, Chasma Nuclear Power Plant unit - 1. 1998, SNERDI.

16. Schulz T. L., Westinghouse AP1000 advanced passive plant. Nuclear Engineering and Design, 2006. 236.

110

17. Demetri K. J. and Saiu G., European Utility Requirements (EUR) Volume 3 Assessment for AP1000. 2004. page(s) 1-7.

18. Makihara Y., et al., Development of the hybrid safety systems for next generation PWR. Progress in Nuclear Energy, 1995. 29(Supplement 1): p. 353-360.

19. Markku H. and Jukka Y., The one-dimensional separate two-phase flow model of APROS. 2008, VTT.

20. Todreas N. E. and Kazmi M.S, Nuclear system I: Thermal Hydraulic Fundamentals. 1990, New York: Hemisphere Publishing Corporation.

21. VTT, Fortum, APROS Thermal Hydraulics: Five Equation Model.

22. Tong, Boiling heat transfer and two-phase flow. 1965: Wiley (New York).

23. Svend T. M., Analysis of the two-fluid model and the drift-flux model fornumerical calculation of two-phase flow, in Department of Energy and Process Engineering. 2006, Norwegian University of Science and Technology.

24. Paul C. and Rafael M., A study of the performance of void fraction correlations used in the context of drift-flux two-phase flow models Nuclear Engineering and Design, 2002. 215(3): p. 199-216.

25. Ishii M. and Hibiki T., Thermo-fluid dynamics of two-phase flow. 2006: Springer.

26. Todreas, N.E., Nuclear system I: Thermal Hydraulic Fundamentals. 1990, New York: Hemisphere Publishing Corporation.

27. Ghiaastiaan S. M., Two-phase flow boiling and condensation. 2008, New York: Cambridge University Press.

28. Motley F. E., Cadek F.F., and Shefcheck J., New Westinghouse Correlation WRB-l for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids. 1977, Westinghouse Electric Corp: Pittsburgh, PA

29. Eveliina T., Modeling of Pressurizer Using APROS and TRACE Thermal Hydraulic Codes. 2006, VTT Research Notes 2339.

30. Groeneveld D.C. and Snoek C.W., A Comprehensive Examination of Heat Transfer Correlations Suitable for Reactor Safety Analysis. Multiphase Science and Technology, 1986. 2: p. 181-274.

31. Bromley L. A., Heat transfer in stable film boiling. Chem. Eng. Prog. Symp, 1950. 46: p. 221-227.

111

32. Puska E. K., APROS Coupling from Core to containment, in M & C,Palais des Papes. 2005: Avignon, France.

33. Heikki P., et al., PACTEL integral test facility – Description of versatile applications. Annals of Nuclear Energy, 2006. 33: p. 994-1009.

34. Virtanena E., Haapalehtoa T., and Kouhiab J., Analysis of steam generator loss-of-feed water experiments with APROS and RELAP5:MOD3.1 computer codes. Nuclear Engineering and Design, 1997. 177: p. 147-153.

35. Jari Tuunanen a, et al., Analyses of PACTEL passive safety injection experimentswith APROS, CATHARE and RELAP5 codes. Nuclear Engineering and Design, 2000. 198: p. 261–286.

36. Tuunanen J., et al., Analyses of PACTEL passive safety injection experiments with APROS, CATHARE and RELAP5 codes. Nuclear Engineering and Design, 2000. 198: p. 261-286.

37. Tuunanena J., et al., Analyses of PACTEL passive safety injection experiments GDE-21 through GDE-25. Nuclear Engineering and Design, 1998. 180: p. 67- 91

38. Porkholm K., et al., apros-based Kola Nuclear Power Plant Analyzer, in 11th Annular Simulator Conference. 1994: San Diego, California.

39. Porkholm K., et al., Loviisa Nuclear Power Plant Analyser,Helsinki. 1992.

40. Joona, et al., thermal hydraulic transient analysis of the high performance light water reactor using APROS and SMABRE, in 20th international conference on structural mechanics in reactor technology (SMiRT 20) ESPOO. August 9-14, 2009 Finland.

41. Hänninen, M., APROS Validation. 2007, VTT, Finland.

42. Geng, Chemical and Volume Control system(manual). 1995, SNERDI.

43. Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors., in 10 volume 1. Revised in 2003, U.S. Government Printing Office: USA. p. Page 735-737.

44. SNERDI, SRH System Manual , SRH System Instrumentation Logic Diagrams. 1994, Shanghai nuclear and design institute.

45. Tahir M., Chughtai I. R., and Aslam M., Qualitative response of safety injection system for a large size break in the reactor coolant system cold leg. Annals of Nuclear Energy, 2007. 34(11): p. 922-925.

46. Yiying, SRH System Manual ,Rev-D. 1995, SNERDI.

112

47. SNERDI, SRC System Manual, SRC System Instrumentation Logic Diagram. 1994, Shanghai nuclear and design institute.

48. SNERDI, SCV System Manual, SCV System Instrumentation Logic Diagram. 1994, Shanghai nuclear and design institute.

49. SNERDI, ―installation drawing of piping and supports SIS NX‖ rev. A. 1994, Shanghai nuclear and design institute.

50. SNERDI, Isometric piping drawing SIS NX rev. A 1994, Shanghai nuclear and design institute.

113

Appendix

Reference power plant simulation

Computer simulation technique provides an alternate environment to observe the response of nuclear systems rather to observe its response on actual power plant. Due to sensitive nature, most of the time, it is not possible to do experimentation on a real power plant. Therefore, it was essential to model reference power plant for providing a platform to our research. This simulation was aimed at representing its complete functionality in reactor safety. Moreover, the model should be capable of replacing its active safety injection system with the proposed passive safety injection system for checking its response in nuclear power plant global environment. For this study a small scale 325 MWe pressurized water reactor has been selected as a reference power plant[15]. This power plant consists of many systems necessary for its working which can be divided in two main categories namely nuclear and conventional islands. Its main parameters of this power plant are listed in Table 6-1. This reactor has thermal power of 998.6 MWt and its corresponding gross electric output is 325 MWe. Since, 25 MWe of power is utilized by the plant systems itself and remaining net electrical output is 300

MWe.

The nuclear steam supply system of reference power plant consists of a reactor coolant system and the associated auxiliary systems necessary for producing steam for driving turbine. The reactor coolant system is designed into two coolant loops connected to the reactor vessel, each containing a reactor coolant pump and a steam generator. To control pressure in the main coolant loop, a pressurizer is attached to hot leg of one of the reactor coolant

114

loops. Highly pressurized coolant moves through the core to remove heat in the core. The heated reactor coolant exits from the reactor vessel hot leg and find its way to the steam generator. In steam generator primary coolant heat is taken up by feed-water and steam is produced. The generated steam in the secondary circuit is used for driving turbine. The electrical generator attached with turbine produced electricity thereupon. The low quality steam is converted to water with the help of a condenser and pumped to steam generator for taking heat in other cycle. In the primary side, the cycle is completed when the coolant from the steam generator tubing side is pumped back to the reactor vessel.

Due to pressurized water in the primary circuit, there are always chances of radioactive water leakage, called LOCA. For controlling the leakage, a number of engineered safety features have been included into the plant design. For this purpose safety systems have been designs. It includes emergency core cooling System, containment spray system and auxiliary feed water system.

All these systems mentioned above are active by design and require an external source for their operation. For this purpose, three types of power sources are available with the plant. It includes standby power supply, batteries and diesel generators. The diesel generators are used as emergency power source. In complete loss of normal off-site power, the diesel generators are available as on-site power supply to the equipment of the engineered safety features. The diesel generators are designed to reach rated voltage, frequency, and able to accept load within 12 seconds, and capable of carrying rated load according to loading sequences within about 60 seconds after receipt of an external intelligent signal called SI signal. The fuel oil tank capacity of each diesel generator train can be operated for a minimum of 14 days under the condition of rated load.

115

Main reactor coolant system simulation

The simulated reactor coolant system performs its functions realistically as per plant requirement. Its main original process flow diagram can be found from final safety report [15]and its simplified process flow diagram is given in Figure A-1. The simulated system is used to transfer heat from reactor core to steam generators, its coolant act as a neutron moderator and reflector.

The simulated major coolant system components are: reactor core, steam generators, pressure vessel, pressurizer relief tank, reactor coolant pumps, pressurizer, safety valves, relief valves and piping.

The control logic has been simulated as per plant requirement as given in the documentation of the reference power plant [47].The brief summary of its control is given below:

The main coolant pumps start manually only when the following conditions are satisfied at the same time:

 Pump seal supply is available when valve SPS-V02A/B open.

 Emergency injection water valve SPS-V06A/B closed.

 HP-cooler bypass valve SPS-V05A/B closed.

 HP-throttle-leakage flow high ≥ 1350 kg/h.

 HP-throttle-leakage flow low ≤ 200 kg/h.

 Injection water seal flow high ≥ 1700 kg/h

 Temperature of injection water after hp-cooler ≤ 52 °C

 Pressure of leakage oil pump (SPS004A/BPO) or leakage oil pump (SPS005A/BPO) ≥ 0.05MPa

116

 Jacking oil pump (SPS03A/BPO) pressure ≥ 1.5 MPa

 Main coolant pumps SRC004ARP/B can be shutdown manually by operator action.

The reactor coolant system pump stops automatically when any one of the following conditions appears:

 Temperature of axial thrust bearing ≥ 110 °C

 Temperature of injection water after high pressure cooler ≥ 57 °C

 Temperature of hp-throttle leakage ≥ 95 °C

 Pressure of injection water before first stage ≤ 2.0 MPa

 Injection water flow ≤ 1500 kg/h in coincidence with emergency injection water valve not open (SPS-V06A/B)

 HP-throttle leakage flow ≤ 200kg/h

 HP-throttle leakage flow ≥ 1350kg/h

 Reactor coolant pump A/B become “on” in 5 seconds in coincidence with reactor coolant pump A/B speed ≤ 200/min

The pressure relief isolating valve SRC-V03A/B will close automatically when pressurizer pressure ≤ 14.7 MPa in coincidence with temperature ≥ 65°C at the downstream of power operated relief valve. However, the valve can be closed manually by operator action when required.

The nitrogen makeup valve SRC-V50 will close automatically when the nitrogen gas pressure in pressure relief tank goes up to 0.022 MPa. The valve will open automatically if pressure drops to 0.081 MPa. However, the valve can be opened /closed manually by operator action when required. The valve SRC-V08 can be opened /closed manually by operator action when required.

117

AUX. SPRAY

~ ~

SPRAY LINE A

SECONDARY

SG A LOOP SG B PRESSURIZER

SPRAY LINE B

LOOP A

REACTOR LOOP B

MCP A MCP MCP B MCP

Figure A- 1 Reactor coolant system simulation diagram

Main charging system simulation

The simulated system performs its functions and provide services to the reactor coolant system realistically as per plant requirement [15]. In this simulation, the primary purpose of this system is charging and maintaining a programmed water level in the pressurizer at its steady state operation. In transient operation it acts as High-high safety injection system and injects borated water in the reactor coolant system. Its simplified flow diagram is given in Figure A-2 and it has the following functions:

 Maintain programmed level in pressurizer and maintain required water inventory in reactor coolant system

 Provides seal water to reactor coolant pumps

118

 Control of reactor coolant water chemistry conditions, activity level, boron concentration and makeup

 Provides filling, draining, and hydrostatic pressure testing of the reactor coolant system

 Emergency core cooling following small LOCA in the reactor coolant system

Major components include: charging pumps, boric acid transfer pumps, regenerative heat exchanger, letdown heat exchanger, excess letdown heat exchanger, seal water reflux heat exchanger, volume control tank, boric acid tanks and letdown orifices.

The simulated model represents control logic as per reference power plant requirement [48].The two centrifugal charging pumps SCV11A/B PO are back up of each other and start automatically when the following signals appear;

 Pressurizer low level in coincidence with at least one of the SCV-235 A/B full open signal

 Pressure at the outlet of SCV-235A/B ≤ 15.7 MPa

 Safety injection signal

When loss of power signal appears, the charging pump will start after the oil pressure of the oil pipe ≥ 0.118 MPa. The centrifugal charging pump stops on following signal; Outlet discharge pressure of pump ≤ 8 MPa, persist for 10 second; when pump start up oil pressure of the oil pipe ≤ 0.078 MPa and when the outlet flow of the charging pump is ≤ 7 m3/h and the SCV-V230A/B have not fully opened in twenty second.

The pump can be manually started or stopped by operator action when required. The oil pumps of centrifugal charging pump are back up of each

119

other. Any one pump can be at „Starting oil pump‟ and other pump will be „standby‟. After receiving the starting signal from the centrifugal charging pump, the starting oil pump will be started. After the centrifugal charging pump running for 90 seconds, the starting or standby oil pump will be stopped automatically. Similarly after receiving the stopping signal from the centrifugal charging pump, the starting oil pump will be started. After the centrifugal charging pump having stopped for 90 seconds, the starting or standby oil pump will be stopped automatically.

The oil pumps (SCV11BPO1/BPO2) of centrifugal charging pump (SCV11BPO) are back up for each other. Any one of the pumps can be as the „Starting oil pump‟ then another pump will be „standby oil pump‟. After receiving the starting signal from the centrifugal charging pump, the starting oil pump will be started. After the centrifugal charging pump running for 90 seconds, the starting oil pump will be stopped automatically. Similarly after receiving the stopping signal from the centrifugal charging pump, the starting oil pump will be started. After the centrifugal charging pump having stopped for 90 seconds, the starting or standby oil pump will be stopped automatically.

The valve SCV-V001/002 will close automatically when pressurizer low-low level signal in coincidence with the full close signals of letdown orifice isolating valves (SCV-V003A/B/C) appear. The valve can be opened /closed manually by operator action when required. The valve SCV-V003A/B/C can be opened when the valves SCV-V001/002 are full opened. The valve cannot be opened when containment isolation signal will appear. The valve closes automatically when pressurizer low-low level signal or containment isolation signal appears. During simulation run, the valve can be closed manually by operator action when required.

The valve SCV-V012 opens automatically when safety injection signal appears. During simulation run the valve can be manually opened /closed by operator action when required. The letdown isolating valve SCV-V015, inside

120

containment, closes automatically when containment isolation signal appears. During simulation run the valve can be manually opened /closed by operator action when required.

The valve SCV-V022 opens automatically when SI signal appears. During simulation run the valve can be manually opened /closed by operator action when required. The containment isolating valve SCV-V035/V041 closes automatically when containment isolation signal appears. During simulation run the valve can be manually opened /closed by operator action when required. The Valve SCV-V042 opens automatically when SI signal appears. During simulation test run the valve can be manually opened /closed by operator action when required. The letdown isolating valve SCV-V043, outside containment, closes automatically when containment isolating signal appears. During simulation run the valve can be manually opened /closed by operator action when required.

The valve SCV-V101 will open automatically when temperature at the outlet of volume control tank ≥ 70C or volume control tank level ≥ 1.75m. The valve automatically closes when volume control tank level ≤ 1.5 m. During simulation run the valve can be manually opened /closed by operator action when required. The outlet isolating valve of VCT SCV-V014A/B will close automatically on the following signal: When VCT-Level ≤ 0.5 m in coincidence with full open signal of valve SCV-V216A/B and SI signal in coincidence with full open signal of valve SCV-V216 A/B. The valve SCV-V104 A/B will open automatically when VCT-Level ≥ 1.5m. When required, the valves can be open and closed manually.

The valve SCV-V126 open automatically, when signal from reactor makeup control system appears. Similarly the valve closes automatically when it receives valve close signal from reactor makeup control system. The valve can be opened /closed manually by operator action when required. The valve SCV-V130 opens automatically when valve open signal from reactor makeup

121

control system appears. Similarly the valve closes automatically when it receives valve close signal from reactor makeup control system. The valve can be opened /closed manually by operator action when required.

The valve SCV-V134 opens automatically when valve open signal from reactor makeup control system appears. Similarly the valve closes automatically when it receives valve close signal from reactor makeup control system. The valve can be opened /closed manually by operator action when required. The SCV- V155 opens automatically when valve open signal from reactor makeup control system appears. Similarly the valve closes automatically when it receives valve close signal from reactor makeup control system. The valve can be opened /closed manually by operator action when required.

The valve SCV-V216A/B opens automatically when volume control level ≤ 0.5 m or the safety injection signal appears. The valve automatically closes when VCT-Level ≥ 1.5m. These valves can be opened /closed manually by operator action when required. The valve SCV-V220 opens automatically when SI signal appears. The valve can be manually opened / closed by operator action when required.

The valve SCV-V222 opens automatically when safety injection signal appears. The valve can be manually opened / closed by operator action when required. The mini-flow isolating valve of centrifugal charging pump SCV- V230A/B will open on auto when flow rate at the outlet of charging pump ≤ 7 m3/h. The valve will close when safety injection signal appears. These valves can be opened /closed manually by operator action when required.

The valve SCV-V234A/B opens automatically when safety injection signal appears. The valve can be manually opened /closed by operator action when required. The valve SCV-V237A/B opens automatically when safety injection signal appears. The valve can be manually opened /closed by operator action when required. The valve SCV-V238A/B opens automatically when safety

122

injection signal appears. The valve can be manually opened /closed by operator action when required. The valve SCV-V407 opens automatically when temperature at the outlet of letdown HX ≥ 50 °C. The valve automatically closes when temperature ≤ 45 °C. The valve can be manually opened /closed by operator action when required.

The boric acid transfer pumps SCV21A/B PO are back up for each other and start automatically on: Start-up signal from makeup control system, working pump failure interlock signal and safety injection signal.

The boric acid transfer pump stops automatically when it receives shutdown signal from the makeup control system. The pump can be manually started /stopped by operator action when required. During simulation test run, the valves SCV-V020/V021, SCV-V028, SCV-V039, SCV-V040, SCV-V154 and SCV- V322 of SCV can be opened manually by operator action when required.

Table A-1 List of main nuclear power plant systems simulated # System 1 Safety Injection System 2 Residual Heat Removal system 3 Reactor Coolant system 5 Main Feed water system 6 Main Steam System 8 Component Cooling System 9 Charging System 10 Auxiliary Feed Water System 11 Pressurizer Liquid Level Control System 12 Pressurizer Pressure Control System 13 Emergency Safety System Actuation 14 Reactor core 15 Essential Service Water System

123

Figure A-2 Main charging system simulation diagram

124

125

Safety injection system simulation

During the simulation of direct injection phase the modeled safety injection system pumps are represented to take suction from refueling water storage tank and inject water to cold legs of reactor coolant system through modeled injection piping. The modeled RHR pumps has been represented to take their suction from refueling water storage tank and provide cooling water to cold legs of reactor coolant system through injection piping under the conditions that the main system pressure would decrease below the shut-off-head of the modeled residual heat removal pumps.

In recirculation phase the safety injection system pumps are boosted by modeled residual heat removal pumps. The residual heat removal pumps take suction from the modeled containment sumps and supply borated water to cold legs of main coolant system and boost the modeled safety injection system pumps as well. Table A-2 through Table A-6 represents a list of safety injection system components simulated, status during normal plant operation, injection mode, recirculation mode and simultaneous recirculation mode.

Table A-2 List of safety injection system components simulated S. no. Equipment Description 1 RWST Refueling Water Storage Tank 2 SIS-02ATK Accumulator A 3 SIS-02BTK Accumulator B 4 SIS01A1PO Safety Injection Pump Group A-1 5 SIS01A2PO Safety Injection Pump Group A-2 6 SIS01B1PO Safety Injection Pump Group B-1 7 SIS01B2PO Safety Injection Pump Group B-2 8 SIS-V02A ACCU.A Discharge Check Valve 9 SIS-V02B ACCU.B Discharge Check Valve

126

10 SIS-V03A ACCU.A Discharge Isolation Valve 11 SIS-V03B ACCU.B Discharge Isolation Valve 12 SIS-V05A ACCU Make Up Water Line Valve 13 SIS-V05B ACCU Make Up Water Line Valve

14 SIS-V10A Accumulator N 2 Isolation Valve

15 SIS-V10B Accumulator N 2 Isolation Valve 16 SIS-V11A Check Valve Leak test Isolation Valve 17 SIS-V11B Check Valve Leak test Isolation Valve 18 SIS-V12A Check Valve Leak test Isolation Valve 19 SIS-V12B Check Valve Leak test Isolation Valve 20 SIS-V13A Test line Containment Isolation Valve 21 SIS-V13B Test line Containment Isolation Valve 22 SIS-V14 ACCU. Make Up Water Check Valve 23 SIS-V15 ACCU. Make-Up Line Valve 24 SIS-V24A Injection Branch Throttle Valve 25 SIS-V24C SIS Injection Branch Throttle Valve 26 SIS-V24D Injection Branch Throttle Valve 27 SIS-V24F SIS Injection Branch Throttle Valve 28 SIS-V24G Injection Branch Throttle Valve 29 SIS-V24I SIS Injection Branch Throttle Valve 30 SIS-V24J Injection Branch Throttle Valve 31 SIS-V24L SIS Injection Branch Throttle Valve 32 SIS-V26A SIS Injection Branch Check Valve 33 SIS-V26B SIS Injection Branch Check Valve 34 SIS-V26C SIS Injection Branch Check Valve 35 SIS-V26D SIS Injection Branch Check Valve 36 SIS-V26E SIS Injection Branch Check Valve 37 SIS-V26F SIS Injection Branch Check Valve 38 SIS-V26G SIS Injection Branch Check Valve

127

39 SIS-V26H SIS Injection Branch Check Valve 40 SIS-V28A SIS Containment Isolation Valve 41 SIS-V28B SIS Containment Isolation Valve 42 SIS-V28C SIS Containment Isolation Valve 43 SIS-V28D SIS Containment Isolation Valve 44 SIS-V28E SIS Containment Isolation Valve 45 SIS-V28F SIS Containment Isolation Valve 46 SIS-V28G SIS Containment Isolation Valve 47 SIS-V28H SIS Containment Isolation Valve 48 SIS-V30 Accumulator Make-up Water Valve 49 SIS-V31A SIS Mini Flow Return Line Valve 50 SIS-V31B SIS Mini Flow Return Line Valve 51 SIS-V32A SIS Pump Mini Flow Check Valve 52 SIS-V32B SIS Pump Mini Flow Check Valve 53 SIS-V32C SIS Pump Mini Flow Check Valve 54 SIS-V32D SIS Pump Mini Flow Check Valve 55 SIS-V33A SIS Pump Discharge Check Valve 56 SIS-V33B SIS Pump Discharge Check Valve 57 SIS-V33C SIS Pump Discharge Check Valve 58 SIS-V33D SIS Pump Discharge Check Valve 59 SIS-V35A SIS Pump Discharge Isolation Valve 60 SIS-V35B SIS Pump Discharge Isolation Valve 61 SIS-V35C SIS Pump Discharge Isolation Valve 62 SIS-V35D SIS Pump Discharge Isolation Valve 63 SIS-V36 RWST Valve from SBR System 64 SIS-V40A SIS Pump Suction Check Valve 65 SIS-V40B SIS Pump Suction Check Valve 66 SIS-V41A SIS Pump Suction Isolation Valve 67 SIS-V41B SIS Pump Suction Isolation Valve

128

68 SIS-V42A Valve from RHR Pump to SIS Pump 69 SIS-V42B RHR Pump to SIS Pump Valve 70 SIS-V43A Valve from RHR Pump to SIS Pump 71 SIS-V43B Valve from RHR Pump to SIS Pump 72 SIS-V59A Sump Discharge Check Valve 73 SIS-V59B Sump Discharge Check Valve 76 SIS-V50A Sump Discharge Line Isolation Valve 77 SIS-V50B Sump Discharge Line Isolation Valve

Table A-3 Safety injection system status during normal plant operation COMPONENT STATUS COMPONENT STATUS SIS01APO1 STAND-BY SISV35A,B,C,D OPEN SIS01APO2 STAND-BY SISV41A,B OPEN SIS01BPO1 STAND-BY SISV28A,C,E,G OPEN SIS01BPO2 STAND-BY SISV28B,D,F,H CLOSE SISV03A,B OPEN SISV15 CLOSE SISV05A,B CLOSE SISV31A,B OPEN SISV36 CLOSE SCV11APO OFF SISV50A,B CLOSE SCV11BPO ON SISV43A,B CLOSE - -

Table A-4 Safety injection system status during injection phase COMPONENT STATUS COMPONENT STATUS SIS01APO1 ON SISV35A,B,C,D OPEN SIS01APO2 ON SISV41A,B OPEN SIS01BPO1 ON SISV28A,C,E,G OPEN SIS01BPO2 ON SISV28B,D,F,H CLOSE

129

SISV03A,B OPEN SISV15 CLOSE SISV05A,B CLOSE SISV31A,B OPEN SISV36 CLOSE SCV11APO ON SISV50A,B CLOSE SCV11BPO ON SISV43A,B CLOSE - -

Table A-5 Safety injection system status during recirculation phase COMPONENT STATUS COMPONENT STATUS SIS01APO1 ON SISV35A,B,C,D OPEN SIS01APO2 ON SISV41A,B CLOSE SIS01BPO1 ON SISV28A,C,E,G OPEN SIS01BPO2 ON SISV28B,D,F,H CLOSE SISV03A,B OPEN SISV15 CLOSE SISV05A,B CLOSE SISV31A,B CLOSE SISV36 CLOSE SCV11APO OFF SISV50A,B OPEN SCV11BPO OFF SISV43A,B OPEN - -

Table A-6 Safety injection system status for simultaneous phase COMPONENT STATUS COMPONENT STATUS SIS01APO1 ON SISV35A,B,C,D OPEN SIS01APO2 ON SISV41A,B CLOSE SIS01BPO1 ON SISV28A,C,E,G valve SIS01BPO2 ON SISV28B,D,F,H switching SISV03A,B - SISV15 CLOSE SISV05A,B CLOSE SISV31A,B CLOSE SISV36 CLOSE SCV11APO OFF SISV50A,B OPEN SCV11BPO OFF SISV43A,B OPEN - -

130

This part of the thesis presents interlocks on safety injection system components, its description and its implementation in simulation. The safety injection system pumps start automatically on safety injection signal. However, during simulation test run, the pumps can be manually started with operator action. These pumps can be shut down by operator action. Interlocks prevent the system from being shutdown until the safety injection signal has been cleared. The safety injection containment isolation valves SIS- V28 A/C/E/G open automatically when safety injection signal appears. However, the valves can be opened manually by operator action when required.

The manual opening of SIS-V28 B/D/F/H valves will be active only when the operator reset it. However, the valves can be closed manually. The isolation valve SIS-V31 A/B of minimum flow line SIS-V31A/B closes automatically when safety injection signal in coincidence with the RWST low low level signal appears (safety injection recirculation signal). The safety injection recirculation signal is memorized when generated. It is cleared only when operator reset it. The manual closing signal of valve will be active only when the operator reset it. The valve can be opened manually. The safety injection pump suction isolation valve SIS-V41A/B will open automatically when safety injection signal appears. The valves close automatically when safety injection signal in coincidence with the refueling water storage tank low-low level appears (safety injection recirculation signal). The safety injection recirculation signal is memorized when generated. It is cleared only when operator reset it. The manual closing signal of valve will be active only when the operator it. However, the valve can be opened manually.

The isolation valve of residual heat removal pump to safety injection pump line SIS-V43A/B will open automatically when safety injection signal in coincidence with the refueling water storage tank low-low level appears (safety injection recirculation signal). The safety injection re-circulation signal

131

is memorized when generated. It is cleared only when operator reset it. The manual open signal of valve will be active only when the operator reset it. However, the valve can be closed manually by operator action when required. The re-circulation sump discharge isolation valve SIS-V50A/B open automatically when safety injection signal in coincidence with the refueling water storage tank low-low level appears (safety injection recirculation signal). The safety injection recirculation signal is memorized when generated. It is cleared only when operator reset it. The manual open signal of valve will be active only when the operator reset it. The valve can be closed manually when required. The manual opening of the valve, SIS- V11A/B will be available only when the operator reset it. However, the valve can be closed manually by operator when required.

The manual opening of the valve, SIS-V12A/B will be available only when the operator reset it. The valve can be closed manually when required. The test line isolation valve SIS-V13A/B will close automatically when containment isolation signal appears. The valve can be opened or closed manually with operator action when required. The accumulator make-up water line containment isolation valve, SIS-V15 closes automatically when containment isolation signal appears. The valve can be opened /closed manually by operator action when required during simulation test run. The valve SIS- V05A/B can be opened or closed manually with operator action when required.

The refueling water storage tank isolation valve SIS-V36 from boron recycle system can be opened or closed manually with operator action when required. The accumulator discharge isolation valve SIS-V03A or SIS-V03B opens automatically when pressurizer pressure is greater than or equal to 7 MPa along with safety injection signal. The valve can be opened manually with operator action. During simulation, the valve can be closed manually by operator action.

132

Simulation of residual heat removal system

The Table A-7 through Table A- 11 represents a list of residual heat removal system components simulated and equipment status during normal plant operation, equipment status during injection mode, equipment status during recirculation mode and equipment status during simultaneous recirculation mode.

Table A-7 Residual heat removal system simulated components list S. No. Equipment Description 1 RHR01APO RHR Pump A 2 RHR01BPO RHR Pump B 3 RHR02AHX RHR Heat exchanger A 4 RHR02BHX RHR Heat exchanger B 5 RHR-V01A Train A Inlet Valve 6 RHR-V01B Train B Inlet Valve 7 RHR-V01C Train A Inlet Valve 8 RHR-V01D Train B Inlet Valve 9 RHR-V02A Cold Leg A Injection Check Valve 10 RHR-V02B Cold Leg B Injection Check Valve 11 RHR-V04A Cold Leg A Injection Throttle Valve 12 RHR-V04B Cold Leg B Injection Throttle Valve 13 RHR-V05A Hot Leg A Injection Throttle Valve 14 RHR-V05B Hot Leg B Injection Throttle Valve 15 RHR-V06A Inlet Isolation Valve With SCV System 16 RHR-V06B Train B Inlet valve With SCV System 17 RHR-V08A Heat Exchanger A Bypass Throttle Valve 18 RHR-V08B Heat exchanger B bypass throttle valve 19 RHR-V09A Heat Exchanger A Outlet Throttle Valve 20 RHR-V09B Heat Exchanger B outlet Throttle Valve

133

21 RHR-V10 RHR HX Outlet To SCV Letdown 22 RHR-V11A Train A Hot leg Injection Check Valve 23 RHR-V11B Train B Hot leg Injection Check Valve 24 RHR-V12A Re-Circulating Line A Throttle Valve 25 RHR-V12B Re-Circulating Line B Throttle Valve 26 RHR-V13A Heat Exchanger Outlet A Isolation Valve 27 RHR-V13B Heat Exchanger Outlet B Isolation Valve 28 RHR-V17A Pump A Outlet Isolation Valve 29 RHR-V17B Pump B Outlet Isolation Valve 30 RHR-V19A Pump A Outlet Check Valve 31 RHR-V19B Pump B Outlet Check Valve 32 RHR-V26A Pump A Inlet Isolation Valve 33 RHR-V26B Pump B Inlet Isolation Valve 34 RHR-V27A RWST To Pump A inlet Check Valve 35 RHR-V27B RWST To Pump B inlet Check Valve 36 RHR-V28A RWST To Pump A Inlet Isolation Valve 37 RHR-V28B RWST To Pump B Inlet Isolation Valve 38 RHR-V29A Train A Inlet Pipe Relief Valve 39 RHR-V29B Train B Inlet Pipe Relief Valve 40 RHR-V30A Cold leg Injection Pipe Relief Valve 41 RHR-V30B Cold leg Injection Pipe Relief Valve 42 RHR-V33A HX A CCW Inlet Isolation Valve 43 RHR-V33B H B CCW Inlet Isolation Valve

Table A-8 Residual heat removal system status during operation COMPONENT STATUS COMPONENT STATUS RHR01APU STAND-BY RHRV28A,B OPEN RHR01BPU STAND-BY RHRV08A,B OPEN

134

RHRV01A,B,C,D CLOSE RHRV09A,B CLOSE RHRV06A, B CLOSE RHRV04A,B OPEN RHRV17A,B CLOSE RHRV05A,B CLOSE RHRV26A,B CLOSE RHRV12A,B OPEN RHRV33A,B CLOSE RHRV13A,B OPEN

Table A-9 Residual heat removal status during injection phase COMPONENT STATUS COMPONENT STATUS RHR01APU ON RHRV28A,B OPEN RHR01BPU ON RHRV08A,B OPEN RHRV01A,B,C,D CLOSE RHRV09A,B CLOSE RHRV06A, B CLOSE RHRV04A,B OPEN RHRV17A,B CLOSE RHRV05A,B CLOSE RHRV26A,B CLOSE RHRV13A,B OPEN RHRV33A,B OPEN RHRV12A,B OPEN

Table A-10 Residual heat removal system during recirculation phase COMPONENT STATUS COMPONENT STATUS RHR01APU ON RHRV28A,B CLOSE RHR01BPU ON RHRV08A,B OPEN RHRV01A,B,C,D CLOSE RHRV09A,B OPEN RHRV06A, B CLOSE RHRV04A,B OPEN RHRV17A,B OPEN RHRV05A,B CLOSE RHRV26A,B CLOSE RHRV13A,B OPEN RHRV33A,B OPEN RHRV12A,B CLOSE

Table A- 11 Residual heat removal system status for simultaneous phase COMPONENT STATUS COMPONENT STATUS RHR01APU ON RHRV28A,B CLOSE RHR01BPU ON RHRV08A,B OPEN RHRV01A,B,C,D OPEN RHRV09A,B OPEN

135

RHRV06A, B CLOSE RHRV04A,B OPEN RHRV17A,B OPEN RHRV05A,B OPEN RHRV26A,B CLOSE RHRV13A,B OPEN RHRV33A,B OPEN RHRV12A,B CLOSE

Residual heat removal system control logic simulation

This part of the thesis presents the description of simulation of residual heat removal system interlocking on its components. The safety injection signal is an indication of accidental condition. During normal simulation run, the pumps of this system start automatically on safety injection signal and stops automatically on the pump shaft seal lost cooling water. However, these pumps can be manually started or stopped by operator action when required. The inlet valves of RHR system are closed automatically when reactor coolant system pressure is greater than or equal to 3.34 MPa. However, these valves can be opened manually only when reactor coolant system pressure has decreased to 2.94 MPa.

The cold leg injection throttle valve (RHR-V04A or RHR-V04B) opens automatically when safety injection signal appears. The valve closes automatically when the following conditions appear simultaneously: Outlet flow of residual heat removal heat exchanger B/A ≥ 50m3/h and outlet flow of residual heat exchanger A or B is greater than the double the outlet flow of residual heat exchanger B/A with safety injection signal. During simulation test run, the valve RHR-V05 A or B can be closed manually by operator action when required. The valve, RHR-V08A or B modeled as per plant requirement and opens automatically on safety injection signal. However, the valve can be closed or opened manually by operator action when required.

The outlet valve of residual heat exchanger, RHR-V09A or B will open manually only when valve close signal of SCS-V14A/B appears. However, the valve can be closed manually by operator action when required. The valve

136

RHR-V10 can be manually opened or closed by operator action when required. The residual heat removal pump mini-flow line valve, RHR-V12A or B opens automatically when inlet flow of residual heat exchanger ≤ 70 m3 /h. The valve closes automatically when residual heat exchanger flow ≥ 120 m3 /h. However, the valve can be opened or closed manually by operator action when required.

The valve RHR-V13A or B opens automatically on safety injection signal. However, the valve can be closed manually by operator action when required. The inlet valve RHR-V17A or B opens manually only when valve close signal of SCS-V15A or B appears. The valve RHR-V26A or B opens automatically on safety injection signal. During simulation run the valve can be opened manually by operator action when required.

During simulation test run in automatic operation the valve RHR-V28A or B closes automatically when open signal of SIS-V50A or B appears. However, the valve can be closed manually at any time during simulation. The valve opens automatically when safety injection signal appears in coincidence with valve, SIS-V50A or B close signal. During simulation the valve RHR-V33 A or B can be opened or closed manually. However, for its automatic operation, the valve opens automatically when one of the following signals appears: Safety injection signal, safety injection recirculation signal and when any one of residual heat removal pumps starts.

Reference power plant simulation data

For incorporation in simulation, the data for different components has been extracted from the documentation of system reference manuals [49, 50]. Particularly for maintain the volume inventory; the piping length of each pipe appearing in the simulation of safety injection system, residual heat removal

137

system and accumulator injection system has been taken and incorporated in the respective system simulation. The detail of this data is available from Table A-12 through Table A-20.

Table A-12 Simulation data for safety injection system pipes Pipe Length Area Design SIM ( m) Design SIM( m2 ) SIS052A-1 35.15100 35.15100 1 323.910 7.25410-02 SIS052A-2 10.40 000 10.40000 323.910 7.25410-02 SIS052A-3 2.300000 2.300000 323.910 7.25410-02 SIS047A-1 3.751500 3.751500 219.18 3.2397410-02 SIS047A-2 10.57800 10.57800 219.18 3.2397410-02 SIS047A-3 3.257500 3.257500 219.18 3.2397410-02 SIS047A-4A 8.078000 8.078000 219.18 3.2397410-02 SIS047A-4B 8.078000 8.078000 219.18 3.2397410-02 SIS046A 2.780000 2.780000 114.35.9 8.2515910-03 SIS046B 2.780000 2.780000 114.35.9 8.2515910-03 SIS036A-1 4.320000 4.320000 88.911 3.515110-03 SIS036A-2 11.00900 11.00900 88.911 3.515110-03 SIS036A-3 12.80000 12.80000 88.911 3.515110-03 SIS036A-4 7.116000 7.116000 88.911 3.515110-03 SIS036B 8.225000 8.225000 88.911 3.515110-03 SIS031A-1 16.85600 16.85600 88.911 3.515110-03 SIS031A-2 3.781000 3.781000 88.911 3.515110-03 SIS031C-1 12.74600 12.74600 88.911 3.515110-03

1 The symbol  represents first entry as pipe inner diameter and second entry as its thickness.

138

SIS031C-2 2.900000 2.900000 88.911 3.515110-03 SIS036C-1 4.320000 4.320000 88.911 3.515110-03 SIS036C-2 11.00900 11.00900 88.911 3.515110-03 SIS036C-3 12.80000 12.80000 88.911 3.515110-03 SIS036C-4 7.116000 7.116000 88.911 3.515110-03 SIS036D 8.225000 8.225000 88.911 3.515110-03 SIS031D-1 16.85600 16.85600 88.911 3.515110-03 SIS031D-2 3.781000 3.781000 88.911 3.515110-03 SIS031F-1 12.74600 12.74600 88.911 3.515110-03 SIS031F-2 2.900000 2.900000 88.911 3.515110-03 SIS052B-1 35.15100 35.15100 323.910 7.25410-02 SIS052B-2 10.40000 10.40000 323.910 7.25410-02 SIS052B-3 2.300000 2.300000 323.910 7.25410-02 SIS047B-1 3.751500 3.751500 219.18 3.2397410-02 SIS047B-2 10.57800 10.57800 219.18 3.2397410-02 SIS047B-3 3.257500 3.257500 219.18 3.2397410-02 SIS047B-4A 8.078000 8.078000 219.18 3.2397410-02 SIS047B-4B 8.078000 8.078000 219.18 3.2397410-02 SIS046C 2.780000 2.780000 114.35.9 8.25210-03 SIS046D 2.780000 2.780000 114.35.9 8.25210-03 SIS036E-1 4.320000 4.320000 88.911 3.515110-03 SIS036E-2 11.00900 11.00900 88.911 3.515110-03 SIS036E-3 12.80000 12.80000 88.911 3.515110-03 SIS036E-4 7.116000 7.116000 88.911 3.515110-03 SIS036F 8.225000 8.225000 88.911 3.515110-03 SIS031G-1 16.85600 16.85600 88.911 3.515110-03 SIS031G-2 3.781000 3.781000 88.911 3.515110-03 SIS031I-1 12.74600 12.74600 88.911 3.515110-03 SIS031I-2 2.900000 2.900000 88.911 3.515110-03

139

SIS036G-1 4.320000 4.320000 88.911 3.515110-03 SIS036G-2 11.00900 11.00900 88.911 3.515110-03 SIS036G-3 12.80000 12.80000 88.911 3.515110-03 SIS036G-4 7.116000 7.116000 88.911 3.515110-03 SIS036H 8.225000 8.225000 88.911 3.515110-03 SIS031J-1 16.85600 16.85600 88.911 3.515110-03 SIS031J-2 3.781000 3.781000 88.911 3.515110-03 SIS031L-1 12.74600 12.74600 88.911 3.515110-03 SIS031L-2 2.900000 2.900000 88.911 3.515110-03 SIS080A 5.367000 5.367000 323.910 7.25410-02 SIS087A 5.004000 5.004000 4578 0.152745 SIS090A 10.11900 10.11900 2738.8 5.12310-02 BSIS080A 5.367000 5.367000 323.910 7.25410-02 BSIS087A 5.004000 5.004000 4578 0.152745 BSIS090A 10.11900 10.11900 2738.8 5.12310-02 SIS048A-1 27.206 19.000 114.35.9 8.2515910-03 SIS048A-2 - 8.20600 114.35.9 8.2515910-03 BSIS048A-1 27.206 19.0000 114.35.9 8.2515910-03 BSIS048A-2 - 8.20600 114.35.9 8.2515910-03 SIS039A-1 4.030500 4.030500 26.94 2.80610-04 SIS039A-3 2.935500 2.935500 26.94 2.80610-04 SIS039B-1 4.030500 4.030500 26.94 2.80610-04 SIS039B-3 2.935500 2.935500 26.94 2.80610-04 SIS039C-1 4.030500 4.030500 26.94 2.80610-04 SIS039C-3 2.935500 2.935500 26.94 2.80610-04 SIS039D-1 4.030500 4.030500 26.94 2.80610-04 SIS039D-3 2.935500 2.935500 26.94 2.80610-04 SIS062 58.552 58.552 168.37.1 1.865110-02

140

Table A-13 Simulation data for safety injection system tanks Component name Volume Length Sim design Sim(m) Design RWST 1695 m3 1695 m3 10.6 10.6 SIS02ATK 56.6 m3 56.6 m3 10.25 10.25 SIS02BTK 56.6 m3 56.6 m3 10.25 10.25

Table A-14 Simulation data for safety injection valves Valve name Area Length Sim ( m2 ) design Sim ( m) design SISV02A 3.698410-02 27328 0.99 0.99 SISV02B 3.698410-02 27328 0.99 0.99

-02 SISV03A 3.698410 273×28 5.187 0.787 SISV03B 3.698410-02 27328 5.187 0.787 SISV05A 2.8110-04 26.94 0.23 0.23 SISV05B 2.8110-04 26.94 0.23 0.23 SISV14 2.8110-04 26.94 5.23 0.23 SISV15 2.8110-04 26.94 5.229 0.229 SISV24A 3.515110-03 88.911 0.343 0.343 SISV24C 3.515110-03 88.911 0.343 0.343 SISV24D 3.515110-03 88.911 0.343 0.343 SISV24F 3.515110-03 88.911 0.343 0.343 SISV24G 3.515110-03 88.911 0.343 0.343 SISV24I 3.515110-03 88.911 0.343 0.343 SISV24J 3.515110-03 88.911 0.343 0.343 SISV24L 3.515110-03 88.911 0.343 0.343 SISV26A 3.515110-03 88.911 0.47 0.47

141

SISV26B 3.515110-03 88.911 0.47 0.47 SISV26C 3.515110-03 88.911 0.47 0.47 SISV26D 3.515110-03 88.911 0.47 0.47 SISV26E 3.515110-03 88.911 0.47 0.47 SISV26F 3.515110-03 88.911 0.47 0.47 SISV26G 3.515110-03 88.911 0.47 0.47 SISV26H 3.515110-03 88.911 0.47 0.47 SISV28A 3.515110-03 88.911 0.305 0.305 SISV28B 3.515110-03 88.911 0.305 0.305 SISV28C 3.515110-03 88.911 0.305 0.305 SISV28D 3.515110-03 88.911 0.305 0.305 SISV28E 3.515110-03 88.911 0.305 0.305 SISV28F 3.515110-03 88.911 0.305 0.305 SISV28G 3.515110-03 88.911 0.305 0.305 SISV28H 3.515110-03 88.911 0.305 0.305 SISV30 2.8110-04 26.94 3.515 0.229 SISV31A 4.71410-03 88.98 0.305 0.305 SISV32A 2.8110-04 26.94 0.229 0.229 SISV32B 2.8110-04 26.94 0.229 0.229 SISV32C 2.8110-04 26.94 0.229 0.229 SISV32D 2.8110-04 26.94 0.229 0.229 SISV33A 3.515110-03 88.911 0.47 0.47 SISV33B 3.515110-03 88.911 0.47 0.47 SISV33C 3.515110-03 88.911 0.47 0.47 SISV33D 3.515110-03 88.911 0.47 0.47 SISV35A 3.515110-03 88.911 0.47 0.47 SISV35B 3.515110-03 88.911 0.47 0.47 SISV35C 3.515110-03 88.911 0.47 0.47 SISV35D 3.515110-03 88.911 0.47 0.47

142

SISV36 2.1482910-03 60.34 0.230 0.230 SISV40A 3.2397410-02 219.18 0.597 0.597 SISV40B 3.2397410-02 219.18 0.597 0.597 SISV41A 3.2397410-02 219.18 0.597 0.597 SISV41B 3.2397410-02 219.18 0.597 0.597 SISV42A 5.345610-03 88.93.2 5.241 0.241 SISV42B 5.345610-03 88.93.2 5.241 0.241 SISV43A 8.2515910-03 114.35.9 5.406 0.406 SISV43B 8.2515910-03 114.35.9 5.406 0.406 SISV59A 7.25410-02 323.910 0.762 0.762 SISV59B 7.25410-02 323.910 0.762 0.762 SISV50A 0.1527 4578 1.16 1.16 SISV50B 0.1527 4578 1.16 1.16

Table A-15 Residual heat removal system simulation data Pipe Length( m ) Area Design sim( m) Design sim ( m2 ) RHR001A 53.527 43.5270 219.122.2 2.39710-02 RHR002A 2.22000 12.2200 219.122.2 2.39710-02 RHR003A-1 15.32700 15.32700 2738.8 5.12310-02 RHR003A-2 0.766000 0.766000 2738.8 5.12310-02 RHR004A-1 17.97900 17.97900 219.18 3.239710-02 RHR004A-2 11.74400 11.74400 219.18 3.239710-02 RHR005A 8.44400 8.444000 219.18 3.239710-02 RHR006A 0.938000 38.7690 168.87.1 1.87710-02 RHR007A 0.693000 3.69300 168.317.5 1.395610-02 RHR008A-1 8.946000 8.94600 2738.8 5.12310-02

143

RHR008A-3 8.583000 8.58300 2738.8 5.123110-02 RHR010A-1 2.073 2.073 168.37.1 1.865110-02 RHR010A-2 11.616 11.616 168.37.1 1.865110-02 RHR011A-1 1.754 1.754 60.34 2.14810-03 RHR011A-2 6.604 6.604 60.34 2.14810-03 RHR013A 2.05400 29.054 60.35.6 1.893510-03 RHR030A 0.5 1.0 33.73.2 5.85410-04 RHR034A 6.5000 11.5000 168.37.1 1.865110-02 RHR035A 34.85900 34.85900 139.76.3 1.26910-02 RHR036A-1 17.23100 17.23100 139.716 9.1110-03 RHR036A-2 2.268000 3.062 139.716 9.1110-03 RHR048A 0.585 1.585 60.34 2.14810-03 RHR055A 21.73000 21.73000 219.122.2 2.39710-02 RHR060A 1.00000 1.00000 2736.3 5.3256410-02 RHR062A 1.00000 1.00000 219.16.3 3.3491110-02 RHR059-1 2.450000 2.450000 60.34 2.14810-03 RHR059-2 22.55000 22.55000 60.34 2.14810-03

Table A-16 Simulation data for residual heat removal pumps Component Area Length name Sim ( m2 ) design Sim ( m ) design RHR01APU 5.123110-02 2738.8 1.0 NA RHR01BPU 5.123110-02 2738.8 1.0 NA

144

Table A-17 Simulation data for accumulator injection system Pipe Length Area design sim(m) Design sim(m2) SIS017-2 07.165 7.165 26.94 2.8110-04 SIS011A-1 00.530 37.478 26.94 2.8110-04 SIS001A 19.914 19.914 27328 3.69810-02 SIS002A 00.870 1.870 27328 3.69810-02 SIS002A-1 - 5.990 27328 3.698410-02 SIS002A-2 02.995 5.000 27328 3.698410-02 SIS002A-3 - 5.000 27328 3.698410-02

Table A-18 Simulation data for safety injection system pumps Component name Area Length Sim(m2) design Sim(m) design (m) SIS01APO1 8.2510-03 114.35.9 0.89 0.89 SIS01APO2 8.2510-03 114.35.9 0.89 0.89 SIS01BPO1 8.2510-03 114.35.9 0.89 0.89 SIS01BPO2 8.2510-03 114.35.9 0.89 0.89

Table A-19 Simulation data for residual heat removal system valves Valve name Area Length sim ( m2) design Sim ( m ) design RHRV01A 2.39710-02 219.122.2 0.762 0.762 RHRV01B 2.39710-02 219.122.2 0.762 0.762 RHRV02A 1.3955610-02 168.317.5 0.914 0.914 RHRV02B 1.3955610-02 168.317.5 0.914 0.914

145

RHRV04A 1.3955610-02 168.317.5 0.762 0.762 RHRV04B 1.3955610-03 168.317.5 0.762 0.762 RHRV05A 9.1110-03 139.716 1.673 0.673 RHRV05B 9.1110-03 139.716 1.673 0.673 RHRV06A 1.8910-03 60.35.6 3.542 0.292 RHRV06B 1.8910-03 60.35.6 3.542 0.292 RHRV08A 1.86510-02 168.37.1 0.5538 0.5538 RHRV08B 1.86510-02 168.37.1 0.5538 0.5538 RHRV09A 3.2397410-02 219.18 0.66 0.66 RHRV09B 3.2397410-02 219.18 0.66 0.66 RHRV10 2.14810-03 60.34 0.292 0.292 RHRV11A 9.1110-03 139.716 1.794 0.794 RHRV11B 9.1110-03 139.716 1.794 0.794 RHRV12A 2.14810-03 60.34 0.292 0.292 RHRV12B 2.14810-03 60.34 0.292 0.292 RHRV13A 1.865110-02 168.37.1 0.6096 0.6096 RHRV13B 1.865110-02 168.37.1 0.6096 0.6096 RHRV17A 3.2397310-02 219.18 0.6605 0.6605 RHRV17B 3.2397310-02 219.18 0.6605 0.6605 RHRV19A 3.2397310-02 219.18 0.597 0.597 RHRV19B 3.2397310-02 219.18 0.597 0.597 RHRV26A 2.3970410-02 219.122.2 0.762 0.762 RHRV26B 2.3970410-02 219.122.2 0.762 0.762 RHRV27A 5.123110-02 2738.8 0.673 0.673 RHRV27B 5.123110-02 2738.8 0.673 0.673 RHRV28A 5.123110-02 2738.8 0.878 0.878 RHRV28B 5.123110-02 2738.8 0.878 0.878 RHRV29A 2.14810-03 60.34 0.315 0.315 RHRV29B 2.14810-03 60.34 0.315 0.315

146

RHRV30A 5.85410-04 33.73.2 1.0 0.5 RHRV30B 5.85410-04 33.73.2 1.0 0.5 RHRV33A 5.325610-02 2736.3 0.33 0.33 RHRV33B 5.325610-02 2736.3 0.33 0.33

Table A-20 Simulation data for residual heat removal heat exchangers Parameters RHR02ahx RHR02bhx Sim design Sim design inside radius of tubes 8.0 mm 8.0 mm 8.0 mm 8.0 mm outside radius of 10.0 mm 10.0 mm 10.0 mm 10.0 mm tubes pipe length for tubes 10.3 m 4616 mm 10.3 m 4616 mm pipe number 724 724 724 724 flow area of shell side 0.68 m2 - 0.68 m2 - length of shell side 2.59 m 3.65 m 2.59 m 3.65 m

147

[Journal Papers]

148

149

150