development of advanced nuclear energy systems in...

International Conference on Advances in Nuclear Science and Engineering in Conjunction with LKSTN 2007 (133-145)

Development of Advanced Nuclear Energy Systems in India

P.D. Krishnani Email: [email protected] Reactor Physics Design Division

Bhabha Atomic Research Centre, Trombay, Mumbai - 400 085, INDIA Abstract

DEVELOPMENT OF ADVANCED NUCLEAR ENERGY SYSTEMS IN INDIA. To ensure long term availability of nuclear energy in a sustainable manner, taking cognisance of its resource position, India has followed the closed fuel cycle and chalked out a three-stage nuclear power programme based on uranium and thorium. The three stages of this programme comprise: (1) Natural uranium fuelled Pressurised Heavy Water Reactors, (2) Fast breeder reactors utilising plutonium based fuel, (3) Advanced nuclear power systems for utilization of thorium. PHWR technology has already reached maturity. The second stage of Indian nuclear power programme aims at setting up of Fast Breeder Reactors for power production and fissile material multiplication. The second stage has started with setting up of a Fast Breeder Test Reactor (FBTR) to understand and master the fast reactor technology. The experience gained in fast reactor technology has led to the development of 500 MWe Prototype Fast Breeder Reactor, the construction of which is going on. The third stage of the Indian nuclear power programme envisages setting up of advanced nuclear power systems based on Thorium-U233 cycle. At present development work is going on the following advanced systems: (1) Advanced Heavy Water Reactor (AHWR); (2) Compact High Temperature Reactor; (3) Multipurpose Nuclear Power Pack; (4) High Temperature Reactor (HTR) for Hydrogen Production. The paper will focus on the salient design features of these advanced systems.

1. Introduction Globally, the abundance of thorium is thrice that of uranium. It is generally agreed that closed fuel cycle and thorium utilisation will be needed for long term sustainability of energy supply in the world. To ensure long term availability of nuclear energy in a sustainable manner, taking cognisance of its resource position, India has followed the closed fuel cycle and chalked out a three-stage nuclear power programme [1] based on uranium and thorium. The three stages of this programme comprise: (1) Natural uranium fuelled Pressurized Heavy Water Reactors, (2) Fast breeder reactors utilising plutonium based fuel, (3) Advanced nuclear power systems for utilization of thorium. Pressurized Heavy Water Reactor (PHWR) is an excellent reactor system for using natural uranium as fuel. The first stage of Indian power programme consists of setting up of a series of PHWRs. There are thirteen operating units of 220 MWe and two units of 540 MWe PHWRs. Three units of 220 MWe PHWRs are under construction and few more units have been planned. PHWR technology has already reached

maturity. The second stage of Indian nuclear power programme aims at setting up of Fast Breeder Reactors for power production and fissile material multiplication. The second stage has started with setting up of a Fast Breeder Test Reactor (FBTR) to understand and master the fast reactor technology. The experience gained in fast reactor technology has led to the development of 500 MWe Prototype Fast Breeder Reactor, the construction of which is going on. The third stage of the Indian nuclear power programme envisages setting up of advanced nuclear power systems based on Thorium-U233 cycle. At present development work is going on the following advanced systems:

• Advanced Heavy Water Reactor (AHWR)

• Compact High Temperature Reactor (CHTR)

• Multipurpose Nuclear Power Pack • High Temperature Reactor (HTR) for

Hydrogen Production AHWR [2] is at very advanced stage of design. It is a 300 MWe, vertical pressure tube type reactor cooled by light water and moderated by

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heavy water. A Compact High Temperature Reactor [3] is being developed as technological demonstration facility for high temperature reactors. It is 100 kWth reactor using uranium-233 as fuel. The design work is going on to develop 600 MWth HTR for hydrogen production and 5 MWth nuclear power pack [4] at remote places not connected to power grid. It may be noted that a number of studies have been performed to use thorium based fuel in existing PHWRs. At present, thoria bundles are used in Indian PHWRs for achieving the initial flux flattening in the core. This represents a unique way of utilizing thorium without any loss of burnup in UO2 fuel. A large number of thoria bundles have been irradiated. The samples from one of these bundles have been analysed for fission products and uranium isotopic composition. These bundles will be used for the development and demonstration of reprocessing of ThO2 bundles in power reactors, and will provide U233 for the development of fabrication technology and irradiation experiments. We shall now discuss the salient design features of above-mentioned advanced systems. 2. Advanced Heavy Water Reactor (AHWR) The main objective for development of AHWR is to demonstrate thorium fuel cycle technologies, along with several other advanced technologies required for next generation reactors, so that these are readily available in time for launching the third stage. The AHWR is a 300 MWe, vertical, pressure tube type, heavy water moderated, boiling light water in natural circulation cooled reactor. The fuel consists of (Th-Pu)O2 and (Th-233U)O2 pins. The fuel cluster is designed to generate maximum energy out of 233U, which is bred in-situ from thorium and has a slightly negative void coefficient of reactivity. For the AHWR, the well-proven pressure tube technology has been adopted and many passive safety features, consistent with the international trend, have been incorporated [5]. The design of the reactor is based on the feedbacks from the extensive analytical and experimental R&D. Important design parameters of AHWR are shown in Table 2.1.

Table 2.1 Important design parameters of AHWR

Reactor power 920 MWth, 300 MWe

Core configuration

Vertical, pressure tube type design

Coolant Boiling light water

Number of coolant channels 452

Pressure tube 120 mm

Lattice pitch 225 mm (square pitch)

No. of pins in fuel cluster

54 [(Th-233U)O2 - 30 pins, (Th-Pu)O2 - 24 pins]

Active fuel length 3.5 m

Total core flow rate 2230 kg/s

Coolant inlet temperature 259 °C (nominal)

Feed water temperature 130 °C

Average steam quality 18.6%

Steam generation rate 414.4 kg/s

Steam drum pressure 70 bar

MHT loop height 39 m

Primary shut down system

37 mechanical shut off rods

Secondary shut down system

Liquid poison injection in moderator

No. of control rods 24

The important design features of AHWR are: (a) Elimination of high-pressure heavy water

coolant resulting in reduction of heavy water leakage losses, and eliminating heavy water recovery system.

(b) Recovery of heat generated in the moderator for feed water heating.

(c) Elimination of major components and equipment such as primary coolant pumps and drive motors, associated control and power supply equipment and corresponding saving of electrical power required to run these pumps.

(d) Shop assembled coolant channels, with features to enable quick replacement of

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pressure tube alone, without affecting other installed channel components.

(e) Replacement of steam generators by simpler

steam drums. (f) Higher steam pressure than in PHWRs. (g) Production of 500 m3/day of demineralised

water in Multi Effect Desalination Plant by using steam from LP Turbine.

(h) 100 year design life of the reactor. The AHWR has several passive safety systems for reactor normal operation, decay heat removal, emergency core cooling, confinement of radioactivity etc. These passive safety features are listed below: • Core heat removal by natural circulation of

coolant during normal operation and shutdown conditions;

• Direct injection of Emergency Core Cooling System (ECCS) water in the fuel Cluster in passive mode during postulated accident conditions like Loss of Coolant Accident (LOCA);

• Containment cooling by passive containment coolers;

• Passive containment isolation of water seal, following a large break LOCA;

• Availability of large inventory of water in Gravity Driven Water Pool (GDWP) at higher elevation inside the containment to facilitate sustenance of core decay heat removal, ECCS injection, containment cooling for at least 72 hours without invoking any active systems or operator action;

• Passive shutdown by poison injection in the moderator, using the system pressure, in case of main heat transfer system high pressure due to failure of wired mechanical shutdown system and liquid poison injection system;

• Passive moderator cooling system to minimise the pressurisation of calandria and release of tritium through cover gas during shutdown and station blackout;

• Passive concrete cooling system for protection of the concrete structure in high-temperature zone.

2.1 Salient features of the physics design of AHWR

The physics design aims at maximising the power from thorium with an external fissile feed in the form of plutonium and hence minimising its consumption. The design envisages maximising the in-situ generation and burning of uranium. This puts an inherent objective of self-sustenance in 233U. The physics safety features include negative core-averaged coolant void reactivity coefficient, negative fuel temperature coefficient, negative power coefficient, low excess reactivity, low speed of control rod withdrawal and low power density. Axial gradation of enrichment is used (the lower half of the fuel assembly is loaded with 4.0% Pu and upper half with 2.5 % Pu in thorium in the outer pins) to achieve better thermal hydraulic margins. Heat removal through natural circulation required that the riser height to be optimised and therefore the heat extracted per channel might be smaller than other reactor types, which increases the core size from the same amount of power to be generated. 2.2 Equilibrium core and approach to

equilibrium core The equilibrium core of AHWR is fuelled by composite clusters consisting of (Th,U)MOX and (Th,Pu)MOX fuel. The composite cluster consists of a circular array of 54 fuel pins [6]. The fuel assembly has a central multi-purpose displacer rod with radial holes for directing ECCS water injection to the fuel pins. The inner and intermediate array of 12 and 18 pins contain (Th, 233U) MOX and the outer 24 pins contain (Th, Pu)MOX. The innermost array of 12 pins has a 233U content of 3.0% by weight and the middle 18 pins have 3.75% 233U. The outer array of (Th, Pu)MOX pins have average of 3.25% by weight of total plutonium. The lower half of the active fuel will have 4.0 % Pu and the upper part will have 2.5 % Pu. The ECCS water enters through the multi purpose displacer region. The displacer region consists of an annular region containing ZrO2. ZrO2 is in the form of powder enclosed in a shell made of SS on the outside and zircaloy on the inside. The fuel cycle is based on the fact the uranium required to fuel the equilibrium core will be bred in-situ and therefore the initial core will have to be fuelled by plutonium and thorium to get the fissile 233U. Thus there will be three phases of operation: initial core, pre equilibrium core and the equilibrium core. In the equilibrium core the

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composite clusters will be loaded, while in the initial core only all (Pu, Th)MOX clusters would be loaded. In the pre equilibrium core a mixture of these two clusters will be employed, based on the gradual availability of composite clusters using the 233U fuel generated in the initial core. 2.3 Core Configuration The AHWR has been designed to produce a total power to the coolant of 920 MWth. The reactor core of AHWR consists of 513 lattice locations in a square lattice pitch of 225 mm. Of these, fuel assemblies occupy 452 locations and 61 locations are reserved for the reactivity control devices and shut down system-1 [6]. Among the 61 locations for the reactivity devices, 37 locations are used for locating Shutoff Rods (SORs). The remaining 24 are used for locating control rods (CRs) for short-term reactivity compensation and power maneuvering during normal operation. The core layout with the locations of the reactivity devices is shown in Fig.2.1. The control element is boron carbide (B C) packed in SS tubes placed between SS shells both for control rods and shut off rods.

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The main heat transport system is designed to transport heat from the fuel through natural circulation of boiling light water coolant. The radial and axial reflector contains heavy water as reflecting material. The reactor has two independent, functionally diverse fast acting shut down systems, namely, shutdown system-1 (SDS-1) consisting of mechanical shut off rods and shutdown system-2 (SDS-2) based on liquid poison injection technology. The reactor is being designed with on-line fuelling capability. 3. Compact High Temperature Reactor (CHTR)

Figure 2.1. Core layout of AHWR With continuously increasing demands, nuclear reactors will need to be increasingly utilized in the future for non-electrical high temperature process heat applications including production of hydrogen or secondary hydrocarbons as a substitute for primary fossil fuel, and for serving as components of compact power packs in remote areas not connected to grid system. Most of the technologies for this need temperatures in the range of 700o C to 900o C. In particular, generation of hydrogen from water using chemico-thermal processes needs high temperatures exceeding 800o C. Keeping this in

mind, a program to design and develop a high temperature reactor system mainly for process heat and non-grid based electricity generation applications has been initiated. Presently, work is going on for designing a Compact High Temperature Reactor (CHTR) of 100 kWth power which would serve as technology demonstration facility. It is being designed on the basis of following guidelines:

• Use of thorium based fuel • Passive core heat removal by natural

circulation of liquid heavy metal coolant

• Passive rejection of entire heat to the atmosphere under accidental condition

• Passive power regulation and shutdown mechanism

• Compact design to minimize weight of the reactor

In the current stage of conceptual design, the

reactor core (Fig. 3.3) consists of nineteen prismatic beryllium oxide moderator blocks. These blocks (Fig. 3.2) contain centrally located graphite tubes (ID/OD 35/75 mm). The central portion of the graphite tube will serve as coolant channel. The liquid metal (Lead-Bismuth eutectic) coolant flows between the top and bottom plenum, upward through the fuel tubes and returning through downcomer tubes. The outlet temperature of coolant is 950 °C. Each graphite tube carries within it the fuel inside 12 equispaced longitudinal bores of 10 mm

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diameter and 700 mm long. These bores will be filled with fuel compacts of approximately 35 mm length made from TRISO particles (Fig. 3.1) embedded in graphite matrix. The TRISO particles are in the form of microspheres of (233U-Th)C2 kernel coated with three layers of soft/hard pyrolitic carbon, SiC and a hard outer carbon layer (90 μm diameter). The fuel, moderator and reflector blocks are contained in a reactor shell made of high temperature and liquid metal corrosion resistant material. The reactor shell is surrounded by two gas gaps that act as insulators during normal reactor operation and reduce heat loss in the radial direction. There is an outer steel shell, surrounded by heat sink. The outer shell has fins to improve heat dissipation. A passive system has been provided to fill the gas gaps with molten metal in case of abnormal rise in coolant

outlet temperature so as to facilitate a conduction path for the reactor heat to outside sink. Nuclear heat from the reactor core is removed passively by a lead-bismuth eutectic alloy coolant, which flows due to natural circulation between the bottom and top plenums, upward through the fuel tubes and returning through the downcomer tubes. On top of the upper plenum, the reactor has multi-layer heat utilisation vessels to provide an interface to systems for high temperature heat applications. A set of sodium heat pipes is in the upper plenum of the reactor to passively transfer heat from the upper plenum to the heat utilisation vessels with a minimum drop of temperature. Another set of heat pipes transfers heat from the upper plenum to the atmospheric air in the case of a postulated accident.

(U+Th)C2 Kernel (250 μm) Pyrolitic Graphite (90 μm) Inner Dense Carbon (30 μm) Silicon Carbide (30 μm) Outer Dense Carbon (50 μm)

Figure 3.1. TRISO fuel particle

Pb-Bi Coolant Graphite TRISO Fuel BeO

Figure 3.2: Fuel Assembly

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Fuel Tubes

Graphite Reflector

PPRS

Downcomers

BeO Moderator

Figure 3.3. CHTR core: schematic description

Table 3.1. Major design and operating characteristics of CHTR

Attributes Design Parameters

Reactor power 100 kW(th)

Core configuration Vertical, prismatic block type

Fuel

233UC2+ ThC2 based TRISO coated fuel particles shaped into fuel compacts with graphite matrix

Weight % of 233U in Th-U 33.75%

Refuelling interval 15 effective full power years Fuel Burnup 68000 MWD/T of heavy metal

Moderator BeO Reflector Partly BeO and partly graphite

Coolant Molten Pb-Bi eutectic alloy (44.5% Pb and 55.5% Bi)

Mode of core heat removal Natural circulation of coolant

Coolant flow rate through core 6.7 kg/s

Coolant inlet temperature 900 °C

Coolant outlet temperature 1000 °C

Loop height 1.4 m (actual length of the fuel tube)

Core diameter 1.27 m

Core height 1.0 m (Height of the fuelled part and axial reflectors)

Primary shutdown system

18 floating annular B4C elements of passive power regulation system

Secondary shutdown system 7 mechanical shut-off rods

For power regulation and control of the reactor, a

to have a long core life to avoid the

h of the 18 control rods surrounding the

passive engineering design feature has been designed to remove the excess reactivity from the system dynamically. In this, the absorber in annular shape floats on lead-bismuth eutectic in liquid form in the center of the 18 hexagonal BeO reflector blocks surrounding the fuel for control and regulation. Sensing the temperature has been chosen for triggering and sustaining control rod movement. A secondary shutdown system consisting of a set of seven shut-off rods of tungsten has been provided, which fall by gravity in the central seven coolant channels. Major design and operating characteristics of CHTR are shown in Table 3.1. t was decidedI

on-site refuelling. This results in large excess reactivity at the beginning of cycle as U233 content is high. To control the initial excess reactivity, a small amount of burnable poison, gadolinium, is added to the central fuel assembly. This configuration gives a core life of about 15 years. The Figure 3.4 shows the effect of the burnable poison.

he wortTfuel assemblies is found to be about 220 mk in hot operating conditions. The maximum worth of a single control rod, when the reactor is in critical state, is about 2 mk. The fuel temperature coefficient for the configuration is calculated as –1.1 × 10–5.

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0 1000 2000 3000 4000 5000 6000 7000 8000

0.96

0.98

1.00

1.02

1.04

1.06

1.08

1.10

1.12

1.14

K-e

ff

Burnup (FPD)

No Poison Gd (40gm)

5500 FPD

0 1000 2000 3000 4000 5000 6000 7000 8000

0.96

0.98

1.00

1.02

1.04

1.06

1.08

1.10

1.12

1.14

K-e

ff

Burnup (FPD)

No Poison Gd (40gm)

5500 FPD

Figure 3.4. Variation of K-eff with burnup

With the above values an analysis was done to assess the effect of an inadvertent withdrawal of a control rod with maximum worth. The analysis showed that even in case of no shutoff-rods falling in, the fuel temperatures are within the permissible limits and power stabilises to a value of about 2-3 times the initial power. Alternative control systems Due to the slow response time of the passive control cum shutdown system (PPRS), it was proposed to have an alternative shut down system. In place of the passive control devices, movement of BeO reflector blocks was studied for control as well as shut-off systems. Two options were considered: axial and radial Movement of the reflector blocks. In both the options, sufficient reactivity margin is found so that any one option can be easily employed for control as well as shut-off purposes. The axial movement of reflector is found to be better in terms of fissile material saving. The proposed configuration is as shown in figure 3.5:

Figure 3.5. Proposed CHTR core layout with alternative control system

Inherent Safety features and passive safety systems CHTR has the following inherent safety features:

i) A strong negative Doppler coefficient of the fuel for any operating condition;

ii) High thermal inertia of the all-ceramic core and low core power density;

iii) A large margin between the normal operating temperature of the fuel (around 1100 °C) and the leak tightness limit of the TRISO coated particle fuel (1600 °C) to retain fission products and gases;

iv) A negative moderator temperature coefficient;

v) Due to the use of the Pb-Bi coolant, which operates at low pressure, there is no over pressurisation and no chance of reactor thermal explosion due to coolant overheating;

vi) Due to a very high boiling point (1670 °C), there is a very large thermal margin to Pb-Bi boiling. This also eliminates the possibility of heat exchange crisis and increases the reliability of heat removal from the core;

vii) There is a negligible thermal energy stored in the coolant and available for release in the event of a leak or accident;

viii) The high temperature Pb-Bi coolant is chemically inert. Even in the eventuality of contact with air or water, it does not react violently with explosions or fires;

ix) No pressure in the coolant allows the use of a graphite coolant channel, improving neutronics of the reactor;

x) A low induced long-lived gamma activity of the coolant; in case of a leakage, the coolant retains iodine and other radio nuclides;

xi) For Pb-Bi coolant, the reactivity effects (void, power, temperature, etc.) are negative.

CHTR employs the following passive systems: i) Natural circulation of coolant to remove

reactor heat during normal operation; ii) Passive regulation of reactor power under

normal operation; iii) Passive shutdown for postulated accidental

conditions; iv) Passive means of conduction of core heat by

filling up the gas gaps with molten metals; v) Passive transfer of reactor heat by heat pipes

under normal and postulated accident conditions;

vi) Passive removal of heat from the reactor core by carbon-carbon composite heat pipes.

4. Multipurpose Nuclear Power Pack The project aims to explore the potential of small reactors without on-site refuelling to supply

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electricity to remote villages, islands, and small-scattered communities in India, which are located in regions not connected to electricity grid. It was also aimed to arrive at a feasible design of small portable nuclear power packs, having long core life, passive safety and reactor core heat removal features, and not needing skilled man-power for operation. To satisfy the energy related needs of the regions mentioned above the unit size of the reactor was estimated to be 5 MWth. A reactor configuration consisting of thorium based metallic fuel, BeO moderator, BeO and graphite reflector, and molten lead alloy based coolant was selected for study. The proposed specification of the reactor is shown in Table 4.1.

Table 4.1. Broad specification of the reactor Attributes Property Reactor power 5 MW(th) Core life Around 10 years Fuel Metallic 233U+ 232Th

+Zr Fuel clad Zircalloy Moderator BeO Reflector material BeO and graphite Coolant Pb-Bi eutectic alloy

1000 mm Core inlet temperature

450 0C

Core outlet temperature

600 0C

No. of fuel assemblies

30

No. of control locations

31

No. of fuel pins per assembly

12

Fuel pin ID 8 mm Fuel pin OD 10 mm Pitch 140 mm Top reflector height 150 mm Bottom reflector height

150 mm

Coolant tube OD 45 mm The core of the reactor is based on tri-angular lattice arrangement. It contains 30 fuel assemblies, 12 in the inner ring and 18 in the outer ring. The cross-sectional view of the reactor core is shown in the Figure 4.1. In each fuel assembly, fuel pins are located in graphite fuel tubes, which also act as coolant tubes. These fuel tubes are located in moderator blocks. These moderator blocks are in turn surrounded by BeO reflector blocks, which

also contain reactor control devices. There are 7 graphite blocks inside the fuel assemblies and 24 BeO reflector blocks outside the fuel assemblies. Graphite reflector blocks surround these BeO reflector blocks. Core height is 100 cm with additional 15-cm top reflector and 15-cm bottom reflector. The proposed fuel for the reactor is metallic fuel with 90% (232Th + 233U) and 10% Zr. Cross-sectional view of a fuel pin is shown in Figure 4.1. The percentage U233 requirement in the alloy for a core life of 3000 FPDs is 14%. Alloy made of these three metals will be made into 8 mm diameter pellet form. These pellets will be encased in a clad of Zircaloy-4 with thickness of 0.75 mm. There will be a radial gap of 0.25 mm between the fuel pellet and the sheath. This gap will be filled with a eutectic alloy consisting of equal amount of Pb, Bi, and Sn. The function of this is to act as a heat transfer medium from the fuel to the clad material.

Proposed Core Configuration

One Fuel Block

Fuel Tube has 12 Nos. of Fuel Pins

One Fuel Pin

Heat Tr. medium (Pb33%+Bi33%+Sn33%)

Metallic Fuel (14%U233 +76%Th +10%Zr)

Clad of Zr-4

Main Reactor ShellOuter Graphite Reflector

BeO ReflectorPassive Power

Regulation System

Inner Graphite Reflector

Fuel CompactsFuel Tube

Pb-Bi Coolant

BeO Moderator

Figure 4.1: Configuration for fuel pin, fuel block and core

Each fuel assembly contains 12 fuel pins arranged in graphite matrix with pitch circle diameter of 65 mm. Central region of fuel assembly has a coolant tube of radius 45 mm.

0 500 1000 1500 2000 2500 3000 3500 40000.8

0.9

1.0

1.1

1.2

1.3

1.4

1.5

With Gd

Without Gd

300 gm Gd in 12 assemblies for 14% U233 in the alloy

K-e

ff

Burn up in FPDs

Figure 4.2: Variation of K-eff with burnup

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As the total control rod worth was found to be much smaller than the total reactivity to be controlled, an effort was made to control the extra reactivity by mixing the burnable poison with the fuel. Different amount of Gd was introduced in different configurations in order to get optimum case with better control and less penalty. It was found that 300 g Gd mixed in the fuel in each of 12 inner fuel assemblies would control the reactivity. Figure 4.2 shows the variation of K-eff with burn up for 2 cases: (1) without Gd and (2) with 300 gm Gd in each of 12 fuel assemblies in the inner fuel ring. With this much Gd, swing in K-eff from cold to hot is negative. The fuel temperature coefficient was also found to be negative. A steady state analysis of the reactor was carried out using finite element method considering conduction heat transfer mode in order to determine the prevalent temperatures in the various components of the reactor. The temperature contours so obtained are shown in Figure 4.3. This reactor incorporates a Passive Power Regulation System (PPRS). These systems can be positioned at the seven inner reflector blocks and/or at 24 outer reflector blocks. This system includes a gas header filled with helium gas at moderate pressure. The header is attached to a driver tube, which contains lead-bismuth eutectic alloy as driven liquid

Figure 4.3: Temperature contours calculated

under steady state conditions The driver tube is housed within a control tube that contains an annular control rod made of boron

carbide with a material compatible to Pb-Bi at that temperature. The annular space between the driver and control tube contains lead-bismuth eutectic, on which the control rod floats. The space above the liquid level is filled with helium. The PPRS gas header, located in the top plenum, is submerged in the coolant and senses the coolant temperature immediately downstream of the heat pipes. Under normal operating conditions, the gas header is surrounded by coolant at 450 °C, the temperature resulting after removal of the reactor power by the heat pipes. Any condition (such as non availability of heat pipes), which causes the coolant to return at a temperature higher than the normal, would also cause the gas in the gas header to heat up. This would lead to a rise in gas pressure in the driver tube and would result in a pressure imbalance between the driver and the control tube. This, in turn, would cause the level of liquid in the driver tube to go down and that in the control tube to go up. Since the absorber rod floats on liquid, it would also rise with the liquid level in the reactor core, thus inserting negative reactivity. Depending on the temperature rise sensed, the system would stabilise at a particular value of reactivity insertion. Cross-sectional layout and schematic diagram of the PPRS are shown in Figure 4.4 and Figure 4.5 respectively. The PPRS operation was analysed using an in-house developed dedicated computer code. A typical analytical result is shown in Figure 4.6. This design would be optimised by changing the process parameters and if required, sizes of the components of the system.

Figure 4.4: Cross-sectional layout of PPRS Figure 4.5: Schematic of PPRS

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Gas Header

Driver tube

Control tube

Pb-Bi eutectic

Absorber

Helium

Absorber

Pb-Bi eutectic

Driver tube

Control tube


0 50 100 150 200 2500

1

2

3

4

5Initial header pressure: 2 AtmDiamter of orifice: 1 mmNo. of orifice: 4

Pow

er (M

Wth

)

Time (Seconds)

Figure 4.6: A typical analytical result

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5. High Temperature Reactor for Hydrogen Production The aim of developing a high temperature reactor is to use nuclear energy to produce hydrogen. Considering very small petroleum reserves and increasing oil prices worldwide, hydrogen can be considered as an alternative to oil for transport applications. The most efficient route for producing hydrogen is Iodine-Sulfur thermo-chemical process for water splitting which needs heat at high temperature (around 850°C). The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. For this reason operating temperature of this reactor was decided to be 1000 0C. Our target is to produce 80,000 Nm3/hr hydrogen, 18 MWe electricity and 375 m3/hr drinking water. In order to meet above requirements reactor power was decided to be 600 MWth. Among various options, Pebble Bed concept was chosen due to its higher level of safety and efficiency. The key feature of this Pebble Bed reactor is the use of coated fuel kernels dispersed in spherical fuel elements (Pebble). Since we are planning to utilize thorium, fuel kernel is (Th232+U233)O2. It is graphite moderated and reflected reactor loaded with randomly packed spherical fuel element and cooled by molten lead. Most of the existing high temperature reactors are gas as coolant. We have decided to use molten lead to achieve natural flow of coolant that move from bottom to top of the core. For this movement we need a chimney. Chimney height reduces with the increase of pebble diameter. In order to reduce chimney height pebble diameter was increased to 10 cm as compared to standard 6 cm. Calculation shows that we can extract 4 kW power from each pebble. So in order to meet full power of 600 MWth total number of pebbles in the core should be 1, 50,000. The proposed design specifications are given in table 5.1.

Table 5.1. Proposed Broad Specifications Reactor power 600 MWth for following

deliverables Hydrogen: 80,000 Nm3/hr Electricity: 18 MWe Drinking water: 375 m3/hr

Moderator Graphite Coolant

outlet/inlet temperature

1000 °C/ 600 °C

Coolant Molten lead Reflector Graphite

Mode of cooling Natural circulation of coolant Fuel 233UO2 & ThO2 based high

burn-up Triso coated particle fuel

Control Passive power regulation and reactor shutdown systems

Energy transfer systems

Intermediate heat exchangers for heat transfer to Helium or other medium for hydrogen production + High efficiency turbo-machinery based electricity generating system + Water desalination system for potable water

Hydrogen production

High efficiency thermo-chemical processes

Description of Core: Reactor core, as shown in Fig. 5.1 has an inner reflector having outer diameter 2 meter followed by reactor core with thickness 1.5 meter and then outer reflector with thickness of 0.5 meter. Core is filled with spherical fuel elements called pebbles with diameter of 10 cm. 59% of the core volume is filled with pebbles and remaining space is occupied by coolant. Description of Pebble: One pebble consists of two radial zones: the inner zone, in which the fissile material in the form of so called Triso particles is uniformly dispersed in the graphite matrix and the outer zone, a shell of pure graphite. Diameter of one pebble in our proposed reactor is 10 cm for the outer zone and 9 cm for the inner zone. Each such pebble contains Triso (tri-isotropic) coated fuel particle, which is a spherical layered composite about 1 mm in diameter. It consists of a kernel of fissile/fertile material surrounded by a porous graphite buffer layer that absorbs radiation damage

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and allows space for fission gases produced during irradiation. Surrounding the buffer layer are a layer of dense pyrolytic carbon, a SiC layer, and a dense outer pyrolytic layer. The pyrolytic carbon layers shrink under irradiation and provide compressive forces that act to protect the SiC layer, which is the primary pressure boundary for the micro-sphere. Challenges in the design: The major challenge in this design is to optimize a pebble configuration with respect to packing fraction of triso- particles in the pebble and percentage of U233 in the fuel kernel so that we can obtain maximum energy with a higher degree of safety. Our target is to obtain a burn up of 900 FPDs in continuous cycle. In order to obtain the required burn up initial K-eff is coming out to be very high. Controlling initial reactivity without facing any burn up penalty is another challenge. Pebble optimization: A parametric study has been done by varying fuel enrichment and packing percentage of triso-particles in a pebble to obtain optimum configuration of a pebble. Based on this study, a packing fraction of 8.6% was chosen for which U233 requirement is 7.3% for a burn up of 900 FPDs. Control of initial reactivity: The reactivity worth of control rods is not sufficient to control initial reactivity. Various options like use of reduced 233U content for the initial core, use of dummy balls or thorium balls are being considered for this purpose. The fuel temperature coefficient of the core has been found to be negative. 6. Acknowledgement Author gratefully acknowledges Umasankari Kannan, Anurag Gupta, Brahmananda Chakraborty and R. Srivenkatesan for their helpful contributions.

ross-sectional view of triso particle and pebble

1. nergy Monograph,

2.

n

3.

n Compact High

4.

ig 5.1: Schematic diagram of reactor core and F

c 7. References

Department of atomic ENuclear Power for National Development: An Indian Perspective, 2001. R.K. Sinha and A. Kakodkar, ‘Design and development of AHWR- The Indian thorium fuelled innovative uclear reactor’, Nucl.

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Engg. And Design, Vol. 236 pp 683-700, 2006. P.P. Kelkar, I.V. Dulera, A. Borgohain, N.K. Maheshwari, R.K. Sinha and P.D. Krishnani, “Conceptual Design Report oTemperature Reactor”, Report No. CHTR/RED/2002/13, 2002. P.D. Krishnani, I.V. Dulera, N.K. Maheshwari, R. Dinesh Babu, B. Chakraborty, A. Gupta, A. Borgohain, A. Basak and P.P. Kelkar, “Design of Multipurpose Nuclear Power Pack for Satisfying Energy Related Needs for Remote Indian Villages”, Progress Report on Second Year’s Work (December

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2005 - November 2006) for the IAEA’s CRP on “Development of Small Reactors Withou

6. Arvind Kumar, Umasankari Kannan, Neelima Prasad Pushpam, B. Krishna Mohan, Anek Kumar, and Amit Thakur ,“Safety Analysis report-Preliminary: Physics Chapter, RPDD / AHWR / 80 / 2007 Rev 0 dated March 29 , (2007).

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On-Site Refuelling”, Vienna, June 4-8, 2007. R.K. Sinha et al, “Design and Development of AHWR – the Indian Thorium Fuelled Innovative Nuclear Reactor”, IT-4, Indian Nuclear Society A

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