Reactor Engineering PDF

Reactor Engineering Fission & Chain Reaction Fissile isotopes namely, 92U233, 92U235 & 94Pu239 atoms undergo fission when their nucleus is bombarded or hit with a neutron 200 MeV heat energy and two or three neutrons are released during fission If these neutrons meet other fissile atoms, they will cause further fission and the process would continue This process is called a chain reaction Uranium containing up to 4% 92U235 is the most common fuel used in nuclear power plants Fission & Chain reaction The process of splitting an atom into two or more fragments is known as fission All atoms do not readily undergo fission Isotopes of elements Uranium (92U233, 92U235 & 92U238), Thorium (90Th232), and Plutonium (94Pu239) are used in the field of nuclear energy Out of these isotopes only 92U233, 92U 235 and 94Pu 239 readily undergo fission They are known as fissile isotopes Isotopes such as 90Th232, 92U238 do not readily undergo fission But, on absorption of a neutron, they get converted into fissile isotopes 92U233, 94Pu239 Such isotopes are known as fertile isotopes Energy from Fission Energy and mass cannot be created or destroyed But they can be changed from one into another Thus in any nuclear reaction, if there is a loss of mass, there has to be corresponding gain in energy E = mc2 Energy From Fission Energy released per fission ~ 200 MeV [~ 32*10-11 J] Mass of U235 required to produce 1 Mega watt day of energy = 108 gm Approximate distribution of energy per fission Ref: Nuclear Reactor Engineering, Samuel Glasstone & Alexander Sesonske Types of Nuclei related to Fission Fissile Nuclides Can undergo fission with neutron of any energy Ex: Uranium 233, Uranium 235, Plutonium 239 Fissionable Nuclides Can undergo fission with sufficiently fast neutrons (> 1 Mev) Ex : Thorium 232, Uranium 238 Fertile Nuclides Can be converted to fissile nuclides Ex : Thorium 232, Uranium 238 Breeding & Conversion Breeder Uses fertile material as blanket and produces more fissile nuclides identical to that used to maintain the fission chain, Eg : U233 fissile material and Th232 blanket Converter or Pseudo breeder - U235 fissile material and U238 blanket Multiplication Factor Effective multiplication factor (Keff) is defined as the ratio of the number of neutrons resulting from fission in each generation to the total number lost by absorption and leakage in the preceding generation Keff = 1 for criticality Four factor formula K pf reproduction factor fast fission factor p resonance escape probability f thermal utilisation factor Keffective pf Pfast Pthermal Four factor formula N Thermal Neutrons Thermal Fissions N pf N fast neutrons K Fast Fissions N N fast neutrons Escape Resonance Capture Resonance Capture N p neutrons N pf thermal neutrons Absorption in non-fuel Absorption in Fuel components Controlled Chain Reaction General features of Nuclear Reactors Reactor Core Moderator Reflector Coolant Control Reactor Classification Classification based on the kinetic energy of the neutrons Thermal reactor -- Majority of fissions is produced by thermal neutrons Fast reactor -- Majority of fissions is produced by fast neutrons Classification based on moderator / coolant Pressurised Heavy Water Reactor (PHWR) Light Water Reactor (LWR) Gas Cooled Reactor Classification based on state of the Pressurised Water Reactor (PWR) Boiling Water Reactor (BWR) Course Name (-01) Different Types of Nuclear Reactors Pressurised Heavy Water Reactor (PHWR) Indian PHWRs (RAPS, MAPS, NAPS, KAPS, KGS, TAPS) CANDUs (Douglas Point, Bruce, Darlington, RAPS ) Pressurised Water Reactor (PWR) VVER KKNPP US-Westinghouse, Framatome -> Areva -> Orano Boiling Water Reactor (BWR) US -Brunswick, Dresden; Japan- Fukushima, TAPS-1&2 Fast breeder Reactor (FBR) UK, France, Germany, Japan, USA, Russia FBTR, PFBR Course Name (-01) Conventional Nuclear Course Name (-01) 24 PRESSURISED WATER REACTOR Course Name (-01) Reactor Assembly BOILING WATER REACTOR Course Name (-01) September 25th to October 25th, 2024 By Orientation Training Program Module-1 of ET-2024 Reactor Engineering Lecture Series September 25th to October 25th, 2024 By Module-1 of ET-2024 September 30th, 2024 Reactor Engineering Lecture No -L2 By Conventional Pressurised Heavy Water Reactor Engineering (NE-04-L2) Reactor 4 Introduction to Indian PHWR design Horizontal reactor vessel Calandria Pressure Tube concept (306/392 channels) Natural Uranium fuelled (Fuel pins; Bundles) Heavy water cooled and moderated Calandria surrounded by water enclosed in a concrete structure Calandria Vault On-power refueling Double containment Suppression Pool Reactor Engineering (NE-04-L2) PHWR SIMPLIFIED FLOW DIAGRAM Reactor Engineering (NE-04-L2) Reactor Assembly Pressure Tube Calandria End-shield End-shield Cold end Hot end Moderator Calandria Tubes Coolant Channel Assembly Reactor Engineering (NE-04-L2) Calandria in PHWR Calandria in Calandria Vault Reactor Engineering (NE-04-L2) Perspective view of PHT system Reactor Engineering (NE-04-L2) Double Containment Primary containment Secondary Major Plant Systems Reactor assembly Control & Shutdown Systems Moderator System Primary Heat Transport System Fuel Handling System Reactor Auxiliaries Process Water Systems Process Control Systems TG & Secondary Cycle Systems Electrical Systems Reactor Engineering (NE-04-L2) Inherent Safety Features of Indian PHWRs Higher neutron generation time Low fissile content Short bundle length limits consequences in case of single bundle failure Passive core cooling feature On power detection of failed fuel Online fuelling and low excess reactivity in the core Moderator as heat sink Reactivity Devices located in low pressure moderator : Rod ejection ruled out Reactor Engineering (NE-04-L2) PHWRs Engineered Safety Features Two independent, diverse and fast-acting Shut Down Systems Emergency Core Cooling through passive accumulators and active pumping system Large volume of sub-cooled moderator acts as heat sink in the case of postulated Loss of Coolant Accident and ECCS failure Additional heat sink available in the form of large volume of water within the calandria vault (Equivalent to core catcher) Double containment Pre-stressed inner containment & RCC outer containment Steel lining for inner containment wall introduced in the latest 700 MWe PHWRs Design checked for capability to withstand small sized military aircraft crash Reactor Engineering (NE-04-L2) Shut Down System # 1 Shut Down System # 2 Flow Schematic Reactor Engineering (NE-04-L2) Fuel cooling provisions NORMAL OPERATION (FISSION HEAT) PRIMARY COOLANT SYSTEM SECONDARY COOLANT SYSTEM (forced circulation) SHUTDOWN (DECAY HEAT) SEVERAL OPTIONS : PRIMARY COOLANT SYSTEM SECONDARY STEAM SYSTEM With circulation by : with SGs fed by any of : Pumps Main BFPs Natural Circulation Aux BFPs Firewater SHUTDOWN COOLING SYSTEM PROCESS WATER Two Loops provided; one is sufficient ACCIDENT CONDITIONS EMERGENCY CORE COOLING SYSTEM ( During LOCA ) MODERATOR HEAVY WATER IN CALANDRIA BACKED UP BY CALANDRIA VAULT WATER Cooling systems essential for safety have power supplies backed by emergency diesel generators Passive Decay Heat Removal System (PDHRS) Schematic PHWR-Engineered Safety Features Reactor Engineering (NE-04-L2) PHWR Features Description PHWR S No A CORE & REACTOR ASSEMBLY 1 Fuel Natural Uranium 2 Moderator Heavy Water 3 Coolant Heavy Water 4 Fuel Housed In Horizontal Pressure Tubes 5 Fuel Geometry · Circular Fuel Bundle · ø 100 mm, 500 mm long Reactor Engineering (NE-04-L2) PHWR Features (Contd) Description PHWR A CORE & REACTOR ASSEMBLY (Contd) 6 Temp Coeff Relativity Small 7 Coolant Void Coeff +ve 8 Flux Tilt Larger Core sizes are prone to flux tilts 9 Reactivity Control Moderator level control/ Rods/ Moderator poison/ Light Water (LZC) Reactor Engineering (NE-04-L2) Description PHWR 10 Shut Down Two fast acting Shut Down systems Rods Liquid Poison based (B / Gd) 11 Flux Ion Chambers, SPNDs Monitoring B REFUELLING 1 Refuelling On-power with fuelling machine clamped on to the coolant channel Spent fuel stored in pool outside RB 2 Failed fuel removal On line Description PHWR S No C REACTOR COOLANT SYSTEM (PHT System) 1 Reactor coolant system Channels/Feeders, isolated from moderator configuration 2 RCS piping · Feeders / Large dia forged headers · Carbon steel Description PHWR D ENGINEERED SAFETY FEATURES Accumulator injection followed by long term recirculation Principle 1 ECCS Accumulator, Suppression pool Source of water Passive+Active Configuration Double containment, Lined inner containment 2 Containment Reactor Engineering (NE-04-L2) S Description PHWR No D ENGINEERED SAFETY FEATURES (Contd) 3 Containment design ~ 14 kg/cm2g pressure 4 Containment pressure Dousing, Suppression pool, spray systems control E ATWS MITIGATION Scheme Two S/D systems, unavailability< 10 per year -6 Reactor Engineering (NE-04-L2) Decay Heat After reactor shut down, substantial amount of heat is generated due to absorption of the energies of the beta particles and gamma rays emitted by the fission products In addition to the fission products decay heat, delay heat is also generated by delayed neutrons emitted by fission products The decay power immediately after reactor shutdown is more than 6% of the reactor thermal power during operation Reactor Engineering (NE-04-L2) Decay Heat Reactor Engineering (NE-04-L2) Indian PHWRs 220 MWe Evolution of Technology Standardisation 540 MWe Scaling up First of a kind systems 700 MWe Integrated Engineering Environment Reactor Engineering (NE-04-L2) Nuclear Power Plants in India (NPCIL) TOTAL PHWRs Capacity Capacity Nos Nos (MWe) (MWe) Operating 22* 6780 18 4460 Projects under 8 6200 6 4200 construction Projects sanctioned 12 9000 10 7000 * RAPS-1 (100 MWe) reactor owned by DAE India has gained expertise and vast experience in design, engineering, construction, operation and maintenance for PHWRs including Fuel fabrication, handling & processing PHWRs contribute to major chunk of Nuclear power in India sksaxena Induction Training Module-1 of ET-2024 September 2024 Reactor Engineering Lecture No -03 () Nuclear Reactor Core In a Reactor Core, nuclear reaction within nuclear fuel generates thermal energy, which is converted to electrical energy Core also contains other in-core components to control the rate of nuclear reaction by moderation, to transfer heat from fuel to outside the reactor core (heat transport system) and structural components to provide structural support to core Reactor core houses:- Fuel Moderator Coolant Reactor Engineering (NE-04-03) PRINCE MISHRA Fundamental Safety Functions Fundamental safety functions to be ensured during all states of a nuclear reactor are:- Control of fission chain reaction (Reactivity Control) Cooling of fuel (Including Decay Heat removal) Confining radioactivity(fuel pellet, clad, Primary Heat Transport (PHT), Inner Containment (IC), Outer Containment (OC), Exclusion Zone) Reactor Engineering (NE-04-03) PRINCE MISHRA Reactor Components Calandria End Shield Coolant Channel Assembly Reactivity Devices Reactor Engineering (NE-04-03) PRINCE MISHRA Calandria (SS 304L) Calandria Key Structure (SS 304L) End Shield (SS 304L) Calandria Vault (CV) (Concrete) Reactor Engineering (NE-04-03) PRINCE MISHRA END SHIELD END SHIELD 37 ELEMENT FUEL BUNDLE Feeder Pipes End Fittings Reactor Face Reactor Engineering (NE-04-03) PRINCE MISHRA CALANDRIA Reactor Engineering (NE-04-03) PRINCE MISHRA INTRODUCTION · Calandria is a horizontal, single-walled, cylindrical, austenitic stainless steel vessel consisting of a main shell, which is stepped down at each end and welded to end shields on either side · Stepped down shells, called small shells, are joined to the main shell by means of flexible annulus plates · Calandria main shell is anchored at four locations in its mid transverse plane to the calandria vault east and west walls by means of four calandria keys two on either side · Each calandria key assembly mainly consists of a female block (keyway), a male block (key) and supporting frame · There are 90 nozzles on top of calandria, for various flux monitoring, regulating, shut down devices, view ports, OPRD & head Tank · There are another 21 nozzles on west side (for HFU, PIU and for Moderator inlet) and 6 nozzles on east side (for moderator inlet) · Four numbers of moderator outlet nozzles are provided at the bottom Reactor Engineering (NE-04-03) PRINCE MISHRA PLAN VIEW OF 700MWe CALANDRIA Reactivity Devices Nozzles Over Pressure Rupture Device Nozzle SIDE VIEW OF Main shell 700 MWe CALANDRIA Calandria Key nulus Plate Horizontal Flux Unit Poison Injection Unit Small shell Moderator Inlet Moderator Outlet Reactor Engineering (NE-04-03) PRINCE MISHRA FRONT VIEW OF 700MWe OPRD NOZZLES (4 Nos) CALANDRIA HEAD TANK NOZZLE H HORIZONTAL C PLANE L DIFFUSER CALANDRIA KEY M OT MANHOLE NOZZLE MODERATOR OUTLET FRONT Reactor Engineering (NE-04-03) PRINCE MISHRA VIEW CALANDRIA Functional Requirements of Calandria Horizontal reactor vessel containing heavy water (D2O) moderator and reflector To house in-core reactivity devices (control, flux monitoring and shutdown devices) To house fuel and coolant channel assemblies (calandria tube, garter spring & coolant tube) Provide inlet and outlet of moderator cooling system To support various piping connected to calandria Provides nozzles for OPRD (Over Pressure Relief Devices) (Rupture discs) Calandria Design Parameters (700 MWe PHWRs) Pressure Loads Design internal pressure = 184 kg/cm2(g) Design external pressure is 128 kg/cm2(g) Design Temperature: 1000C Reactor Engineering (NE-04-03) PRINCE MISHRA Design Parameters Contd Dead weight of Calandria assembly, internals and Reactivity Devices and Drive Mechanisms at top Wt of Cal-40,000 kg; Wt of Mod-263,000 kg; Weight of Reactivity Devices-37,500 kg; Dead weight due to piping-12,000 kg; Buoyancy force on calandria shell-250,000 kg Thermal loads Earthquake loads (OBE, SSE) Piping reactions Sloshing loads CT creep Reactor Engineering (NE-04-03) PRINCE MISHRA Safety & Seismic Classification Calandria is a Safety Class-3 component However it is designed, manufactured and tested as per the requirements of ASME B & PV Code Section III, Division 1, Subsection NB - Class 1 components, for the following reasons: Accessibility is extremely difficult after commissioning of the reactor It forms a boundary between heavy water (moderator & reflector) and the Calandria Vault cooling water (light water) The calandria keys are designed, manufactured, examined and tested as Class 1 component and supports as per ASME B & PV Section-III, Div1, Subsection NF The calandria is a seismic category - 1 component Applicable Codes and Standards ASME- Section III Div1 NB is in general applicable for design, fabrication, examination and testing of the calandria In-service inspection of calandria will be carried out as per requirements of ASME Section XI Division 1 for Class-2 components Section II - Materials Section V - Non-destructive Examination Section IX - Welding and Brazing qualifications Reactor Engineering (NE-04-03) PRINCE MISHRA Calandria Material of Construction MOC is Austenitic Stainless Steel Type 304 L owing to the following advantages, Better corrosion resistant less crud formation High resistance to irradiation damage Low carbon grade to avoid weld sensitization Fast neutron fluence (energy > 1 Mev) on the calandria for 40 years of reactor life is about 45 x 1018 n/cm2 Reactor Engineering (NE-04-03) PRINCE MISHRA S N Components Form Applicable Specification 1 Calandria main shell, Plate ASME SA 240 Type 304 L small shells & annulus plates 2 Nozzles Hollow Forgings ASME SA 182 Grade F 304 L Forged Bars ASME SA 479 Type 304 L 3 OPRD Piping Pipe ASME SA 312 Grade TP 304 L Elbows ASME SA 403 Grade WP 304 L 4 Calandria Keys Plate ASME SA 240 Type 304 L Support Structure 5 Calandria Keys Rectangular ASME SA 182 Grade F 304 L Forging 6 Rupture discs for Reverse buckling As per NPCIL specification OPRDtype Reactor Engineering (NE-04-03) PRINCE MISHRA In 220 MWe PHWRs, there are total 39+23+16=78 Nozzle penetrations from Top, Bottom and Sides of Calandria Top Bottom Sides PSS (Primary Shutdown System)---------- 14 - - SSS (Secondary Shutdown System)-------- 12 12 - RR (Regulating Rods)------------------------ 2 2 SR (Shim Rods)------------------------------- 2 - - AR ---------------------------------------------- 4 4 - In-core Flux Monitoring --------------------- 1 - - Moderator Outlets --------------------------- --------- 4 - Manhole --------------------------------------- --------- 1 - OPRD (Over Pressure Relief Device)------- 4 - - LPIS (Liquid Poison Injection System)----- --------- --------- 2 Moderator inlets 6 nozzles at each side of the calandria Diffusers to distribute the inflow uniformly towards the top of calandria--------- --------- --------- 12 Moderator level measurement ------------ --------- --------- 2 39 23 16 Reactor Engineering (NE-04-03) PRINCE MISHRA 14-Zircaloy Guide Tubes (perforated) for PSS 12-Zircaloy Liquid Shut - off Tubes for SSS Standpipes for ARs, RRs & SR o AR & RR stand pipes are welded at upper and lower end, SR is welded at upper end only o AR (Absorber Rods) & RR (Regulating Rods) are moved in or out of reactor core by means of Flow Tubes o SR (Shim Rods) are moved inside Guide Tubes (perforated) 2 tapping at 5% & 85% of full height for level measurement normal operating level is 96% of full height, rest is filled by cover gas (He) D2O outside Calandria Tube and inside Calandria Shell acts as moderator and reflector for thermalisation of Neutrons The heat produced in moderator due to interaction of neutron is removed by moderator cooling system In 540 / 700 MWe PHWRs, there are total 90+4+27=121 Nozzles penetrations from Top, Bottom and Sides of Calandria Top Bottom Sides SDS#1 (Shutdown System # 1)28 SDS#2 (Shutdown System #2)------------ ---------- ---------- 6 LZCU (Liquid Zone Control Units) 6 AR (Adjuster Rods) 17 CR (control Rods) 4 VP (View Ports) 4 In-core flux monitoring VFU (Vertical Flux Units) 26 HFU (Horizontal Flux Units)--------------- ---------- ---------- 9 Moderator Outlets-------------------------- ---------- 4 Head Tank Nozzle 1 OPRD (Over Pressure Relief Device) 4 Moderator inlets---------------------------- ---------- ---------- 12 90 4 27 Guide Tubes (Perforated) for SDS#1 28-Zircaloy 6-Zircaloy Poison Injection Tubes (Perforated) for SDS#2 Outer Tubes for LZCU 6-Zircaloy 17-Zircaloy Guide Tubes (Perforated) for AR 4-Zircaloy Guide Tubes (Perforated) for CR 26-Zircaloy Outer Tubes for VFU 7 (540 Mwe) / 9 (700 Mwe) Zircaloy Outer Tubes for HFU D2O outside Calandria Tube and inside Calandria Shell acts as moderator and reflector for Thermalisation of Neutrons Supports of Calandria Calandria Supported by means of End shields at North and South ends Calandria K Dead weight - Structure Seismic Load End Shield Axial thermal expansion (constraint) Anchored at four locations to CV by means of calandria key structure Seismic Load (N-S) Allows radial thermal expansion Reactor Engineering (NE-04-03) PRINCE MISHRA Calandria Key Structure 4 Nos of Assemblies Consists of Male Key Block (SA -182 Gr F 304L) Female Key (on Calandria) (SA -182 Gr F 304L) Channels (2 Nos) (SA -182 Gr F 304L) Base Plates (2 Nos) (SA 240 Type 304L) Anchor Bolts (EP) Base Plate Female Key Channel Male Key Provision for Thermal Displacement Thermal Expansion of Gap of 15 mm Calandria Main shell Radial Thermal Expansion Deflection of Annulus plate Deflection of Thermal Expansion of Diaphragm Calandria Small shell Thermal Expansion of 392 Calandria Tubes Thermal Displacement- Axial Thermal Displacement- Radial Reactor Engineering (NE-04-03) PRINCE MISHRA Seismic Loads Plan Calandria Key Calandria E-W Vertical N-S E-W Reactor Engineering (NE-04-03) PRINCE MISHRA Elevation Prince Mishra E-Mail : pmishra September 2024 Lecture No -04 · Each end shield is a cylindrical box closed by the calandria side tube sheet (CSTS) at one end and the fuelling machine side tube sheet (FSTS) at the other end · The box is pierced by 392 lattice tubes arranged on 286 mm square lattice · The space between CSTS, FSTS and end shield main shell, is filled with 10mm diameter carbon steel balls and DM water in the ratio of 57:43 by volume · A shell extension welded to the CSTS of each end shell & the extensions of the end shields are welded to the calandria small shells · The support structure for end shield consists of a cylindrical shell called outer shell, which is concentric with the main shell, and two end flanges welded to form an `H' section along with assembly of saddles with base plate and support plate at the bottom The outer boundaries of end flanges are octagonal in shape Reactor Engineering (NE-04-4) PRINCE MISHRA · The portions of the end flanges and extensions of tube sheets between outer shell and main shell are machined to a thickness of 24mm to serve as flexible diaphragms for accommodating the longitudinal thermal expansion of calandria tubes · One annular shielding ring is provided adjacent to the diaphragm plate on fuelling machine side to enhance shielding capability · Vertical loads are transferred through bottom support structure (assembly of saddles, base plate and support plate) to the calandria vault walls through EP plates and shims · Horizontal loads are transferred through anchor bolts to the calandria vault walls General Arrangement of 700 MWe End Shield(ES) · Water and CS balls are not shown in the figure · CS balls are located only inside inner shell occupying the space between inner shell and lattice tubes · There no CS balls in the annulus space between inner shell and outer shell Reactor Engineering (NE-04-4) PRINCE MISHRA Reactor Engineering (NE-04-4) PRINCE MISHRA 700 MWe ES Reactor Engineering (NE-04-4) PRINCE MISHRA Functions of End Shields Function Facilitated by Provide shielding to limit the dose rate in F/M 1 Carbon Steel balls vault to an acceptable level during shutdown 2 Light water 3 Annular shielding Rings Support and locate coolant channel Lattice Tubes assemblies in which fuel resides and heavy water coolant flows Provide an integral support to reactor 1 Octagonal flanges/ End shield Ring assembly assembly 2 Saddle supports Facilitate removal of heat generated in the Coolable geometry end shield and limit the overall temperature within tolerable limits Provide pressure boundary to moderator and 1 Shell extension/CSTS form a part of calandria vault enclosure 2 Convolution Ring Reactor Engineering (NE-04-4) PRINCE MISHRA SAFETY CLASSIFICATION Class-3 component However, it is designed, manufactured and tested as per the requirements of ASME B & PV Code Section III, Division 1, Subsection NB - Class 1 components, as accessibility is extremely difficult after commissioning of the reactor SEISMIC CATEGORIZATION Seismic Category- 1 component As per AERB guide no: AERB SAFETY GUIDE NO AERB/NPP- PHWR/SG/D-1 for Safety classification and seismic categorization for structures, systems and components of PHWR - End shields including support diaphragms are assigned safety function S, Safety class 3, Seismic category 1 Supports shall meet NF-class 1 Reactor Engineering (NE-04-4) PRINCE MISHRA Description of End Shield Two circular water coolant End Shields of height about 92 m & thickness about 920 mm are located in the north and south Calandria vault Small shells of End Shields and Calandria are welded to each other End Shields attenuate radiation emerging from the reactor core, permitting access to the FM Vaults during reactor shutdown End Shields are supported in opening in the Calandria Vault wall, and form a part of the vault enclosure End Shields are penetrated by 392 Lattice Tubes for Reactor Coolant Channel Assemblies Octagonal frames of Carbon Steel at the outer periphery of End shield are anchored to the Calandria Vault walls with the help of anchor bolts Reactor Engineering (NE-04-4) PRINCE MISHRA Description of End Shield contd In 540 / 700 MWe PHWRs, End Shield is comprised of single compartment packed with Carbon Steel (CS) balls of 10 mm dia and chemistry controlled Demineralised Light Water (57:43 by volume) The DM water in the End Shield is circulated through an external cooling circuit and acts both as neutron shielding and cooling medium Dead weight of Calandria End Shield assembly and its contents is transmitted to the Calandria Vault wall through steel plate pedestals Reactor Engineering (NE-04-4) PRINCE MISHRA Design parameters of ES Design internal pressure: 60 kg/cm2 (g) based on pump shut off head Design external pressure: ES expansion tank is opened to RB atmosphere Therefore no external pressure is considered Design Temperature: 125C This envelopes the expected maximum temperatures of component during all normal operating conditions (Level A) and all upset conditions (Level B) This design temperature limit is chosen based on the following: Leak tight rolled joint of the calandria tube into end shield CSTS Limiting thermal stresses of end shield and limiting the radial expansion of the tube sheets Lowest service temperature (LST): Ambient temperature Design life: The expected design life of end shield is minimum 40 years based on neutron fluence, thermal cycles and corrosion resistance Weight of single end shield excluding loose items - 103 MT and carbon steel balls Weight of loose items like support plate, octagonal - 12 MT locking plate, trunions and lifting lugs Weight of annulus shielding ring - 29 MT Weight of carbon steel balls - 84 MT Sum of above - 228 MT Water hold up in end shield - 17 MT Empty Calandria weight - 40 MT Moderator weight (As calandria vault is filled with - 263 MT water, when calandria is empty, buoyancy will be acted upwards) Weight of Coolant channel and other components - 590 MT (calandria, coolant tube, calandria tube, two numbers of end fittings, fuel bundles, seal and shield plugs, fuel locator, feeder pipes, water holdup, reactivity devices , OPRD) Reactor Engineering (NE-04-4) PRINCE MISHRA Weight transferred through each end shield to - 540 MT concrete structure = weight of end shield + (590MT tons / 2) = { (228 + 17) + 295} Moderator weight not considered as calandria vault is also filled with water Weight transferred through each end shield to - 672 MT concrete structure with moderator weight considered = weight of end shield + (590MT +263MT) / 2) ={ (228 + 17) + 427} Heat load in each ES - 2450 kW Water inlet temp - 515 C Water outlet temp - 56 C Pressure drop in each ES - 2634 kg/cm2 Flow through each ES - 425 cum / hr No of pumps and class of supply - 3 X 50%, Power supply all Class III - Water chemistry pH 85 to 10 conductivity less than 50 micro siemens/cm End Shield Materials of Construction Calandria Side Tube Sheet (CSTS), Fuelling Machine Side Tube Sheet (FSTS), Baffle plate, main shell, shell extension and lattice tubes are made of Austenitic SS- 304 L Austenitic SS- 304 L has Low Nil Ductility Temperature (NDT) Insignificant Shift of NDT with irradiation Reactor Engineering (NE-04-4) PRINCE MISHRA Nuclear requirements for ES & Calandria Nuclear requirements for ES & Calandria are, High resistance to irradiation damage o Corrosion resistance Resistance to radiation embrittlement All these requirements are met by Austenitic Stainless Steel SS-304L Cobalt contents in austenitic SS 304L is restricted to 0025% w/w for ES and 003 % w/w for Calandria Cobalt restriction helps to reduce radiation field in FM vaults and Calandria Vault water (conversion of Cobalt-59 to Cobalt 60 (half life= 52 years)) 220 MWe 540 MWe 700 MWe · Two compartment · Baffle plate · Thickness · Tri-junction joint removed reduced from · Single 1120 mm to compartment 920 mm · Bi-junction joint · Annular · Annular shielding shielding plates ring on both sides on FM side (200 (of 75 mm) mm) Trijunction Weld Dead Weight Saddle Plate Base Support Plate The dead weight is transferred through the tube sheets to the outer shell; from the outer shell to the saddle support and base plate The base plate of end shield is placed on a support plate which is placed on two plate embedded parts (EPs) embedded in CV Reactor Engineering (NE-04-4) PRINCE MISHRA End shield - Calandria Assembly in Calandria Vault Anchor bolts Concret e Grout Locking Plate Octagonal Flange The end shields are placed at the North and South Octagonal opening in the calandria vault The Calandria side octagonal flange and the locking plate are fastened to the anchor bolts embedded in the heavy concrete of CV After installation, the space outside of outer shell is filled with grout Reactor Engineering (NE-04-4) PRINCE MISHRA Provision for Thermal Displacement Asbestos Thermal Expansion of Radial Calandria Main shell clearance Deflection of Annulus plate Thermal Expansion of Deflection of Calandria Small shell Diaphragm Thermal Expansion of 392 Conical Ring allows Radial Calandria Tubes expansion at shell junction Radial Thermal Displacement Radial Thermal displacement is due to thermal Axial Thermal Displacement expansion of tube sheets There are two conical rings Axial thermal displacement is due to the thermal provided at the outside of outer shell to achieve the expansion of calandria shell and calandria tubes shape of grout such that it leaves the space for This is accommodated by the deflection of annulus movement of the tube sheets in the radial direction plate in calandria and the diaphragm plates of ES There is clearance provided in the holes of anchor bolts Reactor Engineering (NE-04-4) PRINCE MISHRA Seismic Loads The vertical and East-West Seismic loads of the reactor assembly is transferred through the end shield outer shell to the grout and to the concrete There is not much of a concern as the end shield in the lateral direction is stiff · The North-South seismic load is shared between the end shields and calandria This is a very critical loading condition for the reactor assembly as the end shield-calandria assembly is flexible in this direction Reactor Engineering (NE-04-4) PRINCE MISHRA Prince Mishra E-Mail : pmishra Orientation Training Programme Module-1 of ET-2022 26th September 2024 Lecture No -05 By () Calandria End shield Coolant Channel Assembly Reactivity Devices Reactor Engineering (NE-04-05) SSK Coolant Channel Assembly Reactor Engineering (NE-04-05) SSK Calandria (SS 304L) Reactor Engineering (NE-04-05) SSK Arrangement of lattice tubes in End shield Reactor Engineering (NE-04-05) SSK Lattice tube CSTS FSTS Reactor Engineering (NE-04-05) SSK TYPICAL PHWR FUEL CHANNEL / COOLANT CHANNEL ASSEMBLY LATTICE TUBE BEARING SLEEVE BEARING SLEEVE (CALSIDE)CSTS FSTS (F/M SIDE) JOURNAL RING (CALSIDE) MODERATOR (F/M SIDE) END FITTING YOKE CALANDRIA GARTER 'E' FACE 'E' FACE TUBE SPRING POSITIONING ASSY STUD PRESSURE END END TUBE FIXED END SHIELD SHIELD FLOATING END COLD GAP (Gap for Creep Growth Expansion) FIG 4231 COOLANT CHANNEL ASSEMBLY - 700 MWe - SCHEMATIC Reactor Engineering (NE-04-05) SSK Reactor Engineering (NE-04-05) SSK L C EAST-WEST VERTICAL CALANDRIA SIDE TUBE SHEET INSERT CALANDRIA TUBE COOLANT TUBE 508 (REF) 1548 (REF) Liner tube EF LINER TUBE ROLLED AREA COOLANT TUBE TO CALANDRIA TUBE SPACER COOLANT TUBE ROLLED AREA FUEL BUNDLE COOLANT CHANNEL ASSEMBLY (DETAILS OF PRESSURE TUBE, CALANDRIA TUBE & GARTER SPRING) PT FIG 4232 SHT 3 OF 3 PT-EF rolled joint Liner tube-EF rolled joint Reactor Engineering (NE-04-05) SSK COOLANT CHANNEL COMPONENTS ARRANGEMENT CALANDRIA GARTER SPRING ANNULUS GAS TUBE PRESSURE TUBE FUEL BUNDLE PHT COOLANT MODERATOR Reactor Engineering (NE-04-05) SSK PHT FEEDER COUPLING LINER TUBE END FITTING POSITIONING ASSEMBLY YOKE (FIXED END) COOLANT CHANNEL ASSEMBLY (DETAILS OF GRAYLOC ASSEMBLY) FIG 4232 SHT 1 OF 3 Reactor Engineering (NE-04-05) SSK COOLANT CHANNEL ASSEMBLY - 3D VIEW Reactor Engineering (NE-04-05) SSK FUNCTION OF COOLANT CHANNEL To house the reactor fuel To direct primary coolant flow Reactor Engineering (NE-04-05) SSK COOLANT CHANNEL ASSEMBLY One coolant channel assembly consists of : Coolant tube/Pressure tube Calandria tube Garter Spring End fitting Assembly Liner Tube Stud and Yoke Positioning Assembly End fitting Bellows Reactor Engineering (NE-04-05) SSK DESCRIPTION OF COOLANT CHANNEL Pressure Tube (PT) is isolated from relatively cold heavy water moderator present in calandria by a concentric zircaloy Calandria Tube (CT) PT attached to End Fitting by zero or low clearance rolled joint thus controlling residual stresses to low level and protecting it against delayed hydride cracking in the long run The PTs are centered and partially supported in the CTs by Garter Spring spacers The annular space between the pressure tube and calandria tube has been sealed by Inconel bellows and is connected to the Annulus Gas Monitoring System (AGMS), which circulates dry carbon dioxide gas + 1% oxygen gas The moisture content of this gas is continuosly monitored at inlet and outlet points to detect possible leaks in the PTs or CTs DESCRIPTION contd Removable Shield Plugs, fitted in the end fittings, provide axial shielding to individual coolant channel PTs are made of Zirconium-25% Niobium alloy low irradiation assisted creep higher strength (lower thickness, hence neutron economy) improved properties with regard to corrosion and hydrogen pick-up Special care is taken in design of coolant channels to ensure that they can be replaced easily when the situation should warrant such a replacement Reactor Engineering (NE-04-05) SSK It has been experienced that garter spring spacers between the calandria tubes and pressure tubes of the earlier design were prone to displacement during operation and hence a modified design of garter spring capable of ensuring their freedom from displacement is used from KAPS, Unit#2 onward reactors The garter spring used in standardized PHWRs are tight fit on the coolant tubes Calandria Tube along with pressure tube/coolant tube are placed in 2286mm square lattice in 220 MWe and 286 mm square lattice in 540 / 700 MWe PHWRs Calandria tube is rolled into calandria side tube sheet at both end by a special sandwich type rolled joint Reactor Engineering (NE-04-05) SSK Coolant Tube/Pressure Tube High pressure & high temperature primary coolant flows through coolant tube Material Selection : Cold worked Zr-25% (wt) Nb Criteria : i) Low thermal neutron absorption c/s ii) Adequate strength iii) Good corrosion resistance under reactor environments iv) Good irradiation damage resistance Disadvantages of Zircaloy-2 over Zr-25 Nb Low strength higher thickness poor neutron economy More hydrogen pickup poor life Reactor Engineering (NE-04-05) SSK Loads acting on coolant tube Internal pressure, Weight of fuel, coolant and self weight, GS reactions Friction at bearing Feeder loads Thermal loads Vibration loads Earthquake loads Operating cycles Allowable stress is Minimum of 1/3 of UTS 2/3 of Ys 67% of avg stress that produce rupture in 100000 hrs 100% of stress that produce creep rate of 10E-7 per hr at 10000 hrs (In PT 1/3 of UTS governs) Coolant Tube/Pressure Tube 220 MWe 130 320 540 MWe 310 125 300 Pressure (MPa) Temperature ( C) 120 290 115 280 110 270 105 260 100 250 0 1000 2000 3000 4000 5000 6000 7000 Distance from inlet end (mm) Variation of design pressure and temperature along the length of pressure tube Reactor Engineering (NE-04-05) SSK Coolant Tube/Pressure Tube Governing Section Governing Section Variation of min required wall thickness along Calandria Tube Function : i) To separate relatively colder moderator from the hotter coolant inside the coolant tube to minimize heat loss ii) To support coolant tube (through Garter Spring) for limiting excessive sag caused by creep Minimum thickness is calculated considering Collapse strength due to external pressure and neutron economy Made of Zircaloy-2 or Zircaloy 4 Seamless Tube expanded at the ends to accommodate laded insert (SS-410) Sandwich rolled joint with CSTS at both the ends Calandria Tube contd Dry CO2 gas is circulated through the annulus space of calandria tube (CT) & pressure tube (PT) AGMS to monitor leak from coolant side or moderator side or from end shield side Dimensions 220 MWe 540 / 700 MWe Coolant tube Calandria tube Coolant tube Calandria tube ID (min) 8255 mm 1077 mm ID (min) 1034 mm 1292 mm Thickness 332 mm 125 mm Thickness 43 mm 14 mm (min) (min) Length 5334 mm 5080 mm Length 6330 mm 6024 mm Calandria Tube Loads acting on CT Garter Spring loads Buoyancy loads External Pressure- cover gas pressure +static head of moderator Tensile force due to pressure acting on Tube Sheet Differential expansion of Calandria and Calandria tubes Annulus Gas Pressure Calandria Tube Design considerations CT should not buckle under external pressure(NOC-08 Kg/cm2) Adjacent CTs may buckle but will not rupture under simultaneous rupture of PT&CT In the event of PT rupture CT is not expected to rupture (Burst pressure>90 Kg/cm2) Garter Spring Spacers Provides intermediate support to the coolant tube and maintain gap between the calandria tube and coolant tube to minimize heat loss from the hot coolant to the moderator Facilitate relative movement between Calandria Tube and Pressure Tube during expansion & contraction of coolant tube Support and transfer a portion of load of pressure tube and its contents and load due to creep sag to calandria tube Tight fit on coolant tubes Total 4 spacers in each channel made up of closed coil helical spring of rectangular c/s (173mm x 104mm) wire and have hooks at both end for tying up Spring Coil diameter is 55 mm to 57 mm They are at distances of 508 mm and 1548 mm from pressure tube center / Reactor vertical centerline on its either side Garter Spring Spacerscontd One girdle ring of round c/s (08mm dia) wire kept inside the garter spring gives additional protection for spring failure and to detect the spring position by Eddy current Garter spring material Zr-25% (wt) Nb 05% Cu material selection criteria : Neutron economy Spring resilience & better strength Corrosion resistance CALANDRIA TUBE SPOT WELDED GIRDLE WIRE PRESSURE TUBE ANGLE OF CONTACT MAXIMUM LOADED TURN OF THE SPRING Fig :- 5 GARTER SPRINGS TIGHT FIT GARTER SPRING Function of End Fitting Assembly a) Transition joint between the coolant tubes and primary circuit piping b) Provides pressure tight connection with F/M during refueling operation without disturbing the coolant flow and provides a thorough path for fuel bundles from F/M to the reactor core c) Supports coolant tube d) Provides shielding for the penetration through the end shield to permit access of F/M vault during S/D condition e) Establish the position of the fuel string in the channel with the help of shield plug f) Supports arrangement for actual creep adjustment Description of End Fitting Assembly Length 2516 mm, Maximum dia 1815 mm and Maximum ID 1375 mm varies along the length (700 MWe) Coolant tubes rolled into the end fitting with 3-groove rolled joints provides adequate pull out strength at maximum operating temperature 300C Material : - Martensitic Stainless Steel (SS-403) forgings in hardened and tempered condition - Hard enough to make a leak proof high pressure joint with F/M head and permit refueling operation A flange for feeder connection on each end of end fitting and feeder is coupled with it by Gray Lock Coupling contd Liner tube inside the End Fitting to guide coolant flow from feeder toward the coolant tube through 247 nos of 16 mm diameter holes provided on its surface Inside diameter : 1039 mm Outside diameter: 11186 mm Length : 21288 mm Shielding plug of 990 mm length is there inside the liner tube and located by a retractable latch which seats in the groove of the liner tube material SS-410 Sealing plug at the F/M end to provide leak proof sealing of the channel when refueling is not there Gap between end fitting and lattice tube is stepped down to prevent radiation streaming Description of Stud and Yoke Positioning Assembly Consists of split yoke assembly, studs, nuts and lock nuts Studs threaded into the Fueling-Machine Side Tube Sheet (FSTS) nuts mounted on the studs adjusted periodically to accommodate coolant tube periodical elongation Material of yoke assembly SS 304 Function of Positioning Assembly i) Postulated event of pressure tube rupture coolant channel debris with end fitting moves with high speed towards F/M side to prevent it some shock absorbing system is required which will absorb the kinetic energy of the coolant channel component ii) to provide accommodation and adjustment for longitudinal creep FUELLING MACHINE SIDE TUBE SHEET LATTICE TUBE RETAINING RING FOR LATTICE TUBE BEARING SLEEVE (F/M SIDE) (F/M SIDE) END FITTING BELLOWS ASSEMBLY JOURNAL RING (F/M SIDE) POSITIONING ASSEMBLY YOKE (FIXED END) FOR FLOATING END SEE DETAIL 'X' TO BE JAMMED AGAINST END OF THREADS F/MSIDE CAL SIDE DETAIL 'X' - YOKE ASSEMBLY (FLOATING END) COOLANT CHANNEL ASSEMBLY (DETAILS OF END FITTING, SEAL BELLOWS & POSITIONING ASSEMBLY) FIG 4232 Reactor Engineering (NE-04-05) SSK SHT 2 OF 3 2 Halves A & B POSITIONING ASSEMBLY - SPLIT YOKE (A & B) 1 PRESSURE TUBE 12 LATTICE TUBE F/M SIDE 2 CALANDRIA TUBE RETAINING RING 16 15 YOKE 3 13 END FITTING 3 GARTER SPRING 16 CSTS 4 END FITTING RETAINING RING 13 9 17 FSTS 14 STUD 21 5 LINER TUBE 7 11 1 6 END FITTING BELLOWS 18 FUEL LOCATOR 1 2 7 BEARING SLEEVES (CAL SIDE) 19 SEAL PLUG 17 29 8 BEARING SLEEVES (F/M SIDE) 20 SHIELD PLUG 23 8 9 JOURNAL RINGS (CAL SIDE) 24 21 FUEL BUNDLE 28 4 10 JOURNAL RINGS (F/M SIDE) 22 HPFC 11 LATTICE TUBE CAL SIDE 18 L CA 23 PT ROLLED AREA 10 RETAINING RING E SID 12 6 24 LT ROLLED AREA 20 25 FRONT NUT 27 22 14 90 ) 27 26 LOCK NUT M (NO M F/ 15 DE 26 FOR FLOATING END SI 5 26 27 BACK NUT 25 SEE DETAIL `C' DETAIL 'C' 25 28 BELLOW PROTECTIVE SLEEVE ALL DIMENSIONS ARE IN MM 19 29 LANDED SLEEVE INSERT Reactor Engineering (NE-04-05) SSK End Fitting Bellows Functions : i) Provides sealing to the annulus space of pressure tube and calandria tube ii) Accommodates relative displacements between the end shields and coolant channels iii)Provides inlet outlet connection to AGMS in 220 MWe, whereas in 540 / 700 MWe it is on the Lattice Tube Material Inconel - 600 Brahm Prakash Contact Nos 25994618 (Off), Cell : 9969339559 E-Mail : braham SS Khandave Contact Nos 25994316 (Off), Cell : 9869618169 E-Mail : sskhandave Reactor Engineering (NE-04-05) SSK Module-1 of ET-2024 Reactor Engineering Lecture No -06 A () Introduction to Shielding Shielding is a physical entity interposed between a source of ionizing radiation and an object to be protected During reactor plant operation, the shield design of the reactor system should be such that the working personnel receives minimum dose as per standards set for working personnel, working in radiation zone Neutron and gamma radiations which are attenuated by the medium deposit their energy in the form of heat in the medium and shield design must take this factor also into account Reactor Engineering (NE-04-6A) Introduction to Shielding contd The impinging neutron on structural materials like carbon steel may displace the atoms from their normal position This leads to reduction in the strength of the material; it loses its physical and mechanical properties Neutron fluence on a structure is kept in limits to avoid the radiation damage During (n, ) reaction the nuclei become radioactive and may become cause of concern if the area near the material which have been activated has to be approached For example in steel nickel is present and with it 04% Co-59 isotope impurity is present During reactor operation due to Co59(n,)Co60, radioactive nuclide Co-60 is produced which emits 2 gamma photons of energy 117MeV and 133MeV So material activation have also to be taken into consideration Reactor Engineering (NE-04-6A) Aspects of reactor shielding Three principles of shielding the reactor are as follows: Slowing down of fast neutrons : elastic / inelastic scattering low Z / moderately high Z Capture of the slowed down (or initially slow) neutrons: water, iron, boron etc Attenuation of all forms of Gamma radiations: high Z materials Reactor Engineering (NE-04-6A) Types of shielding materials Materials used in shielding can be divided into three broad categories based on their function: Heavy / moderately heavy elements : to attenuate gamma & to slow down fast neutrons to about 1 MeV by inelastic collisions Hydrogenous substances : moderate neutrons having energies in the range below ~ 1 MeV by elastic collisions Materials which capture neutrons : eg materials containing Boron Reactor Engineering (NE-04-6A) Common shielding materials Light Water : Hydrogenous ; elastic scattering Steel : Inelastic scattering ; capture of slowed down neutrons Concrete : Neutrons & Gamma Lead : High Z; for Gamma; Low Melting Point Boron containing materials : Boral, Boron impregnated wood, borated steel etc Reactor Engineering (NE-04-6A) Radiation Streaming through shield During mechanical design, sometime clearances are left between the penetrations and devices for proper installation and operation Through these clearances radiation streaming takes place and this gives rise to radiation level locally at the site of these penetrations So radiation streaming analysis is done to estimate the effect of these holes or penetrations in the design Calandria (SS 304L) End Shield (SS 304L) Calandria Vault (CV) (Concrete) Reactor Engineering (NE-04-6A) Calandria Vault Shielding and Sealing Calandria closed and supported at each end by End Shields is located in light water filled, steel lined, concrete vault which provides thermal shield Calandria Vault comprises of rectangular reinforced concrete structure, closed at the top by heavy density concretefilled stiffened Steel Box known as Top Hatch Beam with steel Deck Plate Assembly above the Top Hatch Beam for further shielding Top Hatch and Deck Plate have penetrations for supporting the Vertical Reactivity Devices accurately Concrete is not for strength but for attenuating Gamma & Neutron radiation Sealing of the Calandria Vault is done by means of Sealing Bellows attached between the penetration in Vault and Top hatch Beam and Vertical and Horizontal Reactivity Devices Reactor Engineering (NE-04-6A) Reactor Engineering (NE-04-6A) HFU EP Reactor Engineering (NE-04-6A) HFU EPs 6 Nos EQUISPACED TAPPED HOLES ON 2840 PCD 40 M 24 x 30 (TYP) (1683 OD) SQ 190 150 NB ID OF 170 SQ 300 SQ 200 = Ø 230 500 Ø 40 (TYP) 40 X 10 THK 10 THK 745 495 10 THK 25 20 x 45° (TYP) 60 (TYP) 1500 1000 350 = = 2556 (REF) Two tier arrangement for Reactivity Devices over calandria vault top hatch beam Reactor Engineering (NE-04-6A) Renjith Kumar M / Satyartha Singh Contact Nos2599 4320/4342 (Off) E-Mail : renjithkm satyarth Reactor Engineering (NE-04-6A) Module-1 of ET-2023 Lecture No -06B () OUTLINE What is Reactivity ? Multiplication Factor & Reactivity Reactivity & Reactor Power Importance of Delayed Neutrons Why control Reactivity ? Methods of Reactivity Control Location of Reactivity Devices Absorber Materials Types of Reactivity Devices Design Manufacturing Requirements Functions of Reactivity Devices Reactor Engineering (NE-04-06B) RKM / SS WHAT IS REACTIVITY? A measure of a reactor's departure from criticality Predict how the neutron population of a reactor will change over time A measure of the imbalance between neutron production and loss across the entire core Reactor Engineering (NE-04-06B) RKM / SS MULTIPLICATION FACTOR & REACTIVITY Effective multiplication factor (Keff) : ratio of the number of neutrons resulting from fission in each generation to the total number lost by absorption and leakage in the preceding Reactivity, =(keff -1)/ keff · Reactivity is a dimensionless number ·Keff < 1 implies < 0, · Commonly used units are mk, pcm (percent millirho), dollars etc · Keff = 1 implies = 0 · mk = 0001 k (Keff = 1 for criticality), and · 1 pcm = 000001 k/k · Keff > 1 implies > 0 · Reactivity in Dollars = ,where, = delayed neutron fraction which is 00065 for U-235 eg If K = 1007, =7 mk Reactor Engineering (NE-04-06B) RKM / SS REACTIVITY & REACTOR POWER Reactivity is directly related to the tendency of the reactor core to change power level: if is +ve, the core power tends to increase if is -ve, the core power tends to decrease if is zero, the core power tends to remain same The reactivity of the core is adjusted by the reactor control system in order to obtain a desired power level change (or to keep the same power level) Reactor Engineering (NE-04-06B) RKM / SS IMPORTANCE OF DELAYED NEUTRONS Delayed neutrons are produced by radioactive decay of fission fragments Daughter Nucleus Daughter Nucleus decay emits Fission Product (excited state) (stable) neutron These neutrons are `delayed' as their emission is determined by the rate of decay of the precursor fission product Delay in release of a proportion of fission neutrons causes an increase in their average effective lifetime Effect of delayed neutrons is to prevent the neutron density from rising rapidly Were it not for the presence of delayed neutrons, the control rods could not be moved in and out of the reactor with sufficient speed to control the reactor power Delayed neutrons make controlled nuclear fission technically possible Reactor Engineering (NE-04-06B) RKM / SS WHY CONTROL REACTIVITY ? To operate a nuclear reactor at a steady power, keep Keff = 1 ( = 0) counteract perturbations flux flattening for uniform fuel bundle power output add or remove neutron absorber To reduce power or shut down the reactor, make Keff < 1 insert negative reactivity done by adding neutron absorber While starting up the reactor or when increase in power is required, make Keff slightly greater than 1 insert positive reactivity for a short time done by removing neutron absorber Reactor Engineering (NE-04-06B) RKM / SS METHODS OF REACTIVITY CONTROL Inserting or removing neutron absorbing material cadmium, boron, cobalt, stainless steel etc reduce or increase the thermal utilization factor (f) and thus changing Keff Addition or removal of fissile material Changing neutron leakage from the reactor By increasing or decrease core size (achieved by increasing or decreasing moderator level to cover more or less fuel) By increasing or decreasing reflector thickness (again by moderator level control) Reactor Engineering (NE-04-06B) RKM / SS LOCATION OF REACTIVITY DEVICES The effectiveness, or worth, of a control rod depends largely upon the value of the neutron flux at the location of the rod Worth is roughly proportional to the square of the neutron flux at that location before rod insertion The control rod will have maximum effect if it is placed in the reactor where the flux is a maximum In PHWRs, all in-core reactivity devices are located in the low- pressure moderator environment, interstitially between the rows of coolant channels ABSORBER MATERIALS In PHWRs: Cadmium is used as a neutron absorbing material in Shut off /Control Rods in PHWRs Gadolinium & Boron are used as liquid poison Stainless Steel, Cobalt & Light Water as also used as absorber materials Cadmium is not used in PWRs since water outlet temperature in PWRs is about 330oC which is close to melting point of Cd (321oC) In PWRs alloys of Hafnium, Silver Indium Cadmium alloy and Boron containing substance are commonly used in Control Rods TYPES OF REACTIVITY DEVICES DESIGN Mechanical devices in the form of rods, moving in designated guide tubes inside the core Magnetic Jack type arrangement for Control Rods in PWRs Rope & Sheave Drive Mechanism for Shut Off Rods, Control Rods & Adjuster Rods in 540/700 MWe PHWRs Liquid poison mixed in coolant / moderator or contained in tubes with provisions to change their levels Boron addition in PWRs Liquid Zone Control units having compartments filled with light water in 540/700 MWe PHWRs Gadolinium Nitrate mixed with moderator as Shut Down System #2 in Lithium Pentaborate filled in Shut of tubes as Secondary Shut Down System in 220 MWe PHWRs MANUFACTURING REQUIREMENTS Cd is sandwiched between SS Tubes since it cannot be directly used as neutron absorber tubes because of its low mechanical strength Silver-Indium-Cadmium alloy and Boron Control Rod materials are enclosed in SS tube to protect them from corrosion by high temperature water Hafnium is resistant to corrosion by high temperature water, it can be readily fabricated and has adequate mechanical strength Reactor Engineering (NE-04-06B) RKM / SS FUNCTIONS OF REACTIVITY DEVICES 1) Reactor Regulation:- maintain keff = 1 for steady power operation provide small change +ve or ve in k prevent development of flux oscillations 2) Reactor Protection:- rapidly insert large amount of negative reactivity to shut- down the reactor (TRIP) 3) Flux Monitoring:- in-core and ex-core neutron flux measurements Reactor Engineering (NE-04-07A) RKM Renjith Kumar M / Satyartha Singh Contact Nos 022-25994320/4342 (Off) E-Mail : renjithkm satyarth Reactor Engineering (NE-04-06B) RKM / SS Lecture No -07 & 08 LIQUID ZONE CONTROL SYSTEM ADJUSTER ROD REGULATION MECHANISM CONTROL ROD SHUT OFF ROD (SDS#1) REACTIVITY SHUT DOWN DEVICES POISON INJECTION SYSTEM (SDS#2) VERTICAL FLUX UNIT IN-CORE HORIZONTAL FLUX FLUX UNIT MONITORING IONIZATION EX-CORE CHAMBER ASSEMBLY Reactor Engineering (NE-04-07A) RKM / SS METHODS OF REACTOR REGULATION Moderator Level Control (+ or ) a) Changes of moderator level changes the reflector on top of reactor thus varying Leakage (f & t) Advantages: Easily incorporated into a system using moderator dump for protection Disadvantages: Zone control is not possible Lowering the moderator level distorts the overall flux distribution b) Liquid Poison Injection System ( ) Long term reactivity control use soluble poison to moderator Lesser flux distortion compared to solid absorber rod Eg: Gadolinium in the form of Gadolinium Nitrate Gd (NO3)3 6H2O and Boron in the form of Boric Acid D3BO3 Reactor Engineering (NE-04-07A) RKM / SS METHODS OF REACTOR REGULATION c) Liquid Zone Control (LZC) (+ or ) LZC inside the reactor core varies the light water (mild neutron absorber) level based on the zonal and bulk power requirements Advantages: 1) Individual zone levels can be independently varied for zone control 2) Operating equipment is accessible during reactor operation 3) Cooling easily accomplished 4) Only slight distortion of the overall flux pattern Disadvantages: 1) Requires special design to ensure that the zones fail safe (ie, fill) 2) In core structure represents a reactivity (or fuel burn up) loss Reactor Engineering (NE-04-07A) RKM / SS d) Adjuster Rod (+ ) Parasitic Neutron Absorber Stainless Steel or Cobalt normally fully inserted Located vertically between Calandria Tubes decreases the thermal utilization factor f Positive reactivity added by withdrawing the ARs Provide flux flattening and xenon override No significant decrease in reactivity worth over normal lifetime Presence of adjusters results in a fuel burnup penalty of -8% Reactor Engineering (NE-04-07A) RKM / SS e) Control Absorber ( ) Hollow Cylinder eg Parasitic Neutron Absorber - Cadmium sandwiched between Stainless Steel tubes operated vertically between Calandria Tubes change the thermal utilization factor f Provide additional reactivity at minimal cost compared to LZC Disadvantage: In-core guide tubes represent, permanent, reactivity loss (fuel burn up loss) Reactor Engineering (NE-04-07A) RKM / SS Locations of Vertical Reactivity Devices Reactor Engineering (NE-04-07A) RKM / SS LIQUID ZONE CONTROL UNITS Reactor Engineering (NE-04-07A) RKM / SS Core Diameter and Reflector thickness in 540/ 700 MWe PHWR core Calandria Tube p p p Calandria Shell Here, p = pitch ie center to center distance between two Calandria Tubes D n = No of Calandria Tubes D = Calandria Internal diameter Dc = Core diameter t = Reflector thickness Reactor Engineering (NE-04-07A) RKM / SS WHY LIQUID ZONE CONTROL ? As the reactor core size increases, the Migration length becomes a smaller fraction of the core dimensions each neutron zone of influence becomes a smaller fraction of the core volume various core regions becomes less tightly coupled or loosely coupled Core Diameter to Migration length ratio for 540 / 700 MWe PHWRs works out to 640 cm/20 cm = 32 whereas for 220 MWe it is 450/20 = 225 Since the D/M is greater than 25; Zonal or Spatial control is necessary Xenon induced oscillations which is a fundamental instability present due to loosely coupled core is suppressed by LZC System LZC System is important in 540 / 700 MWe PHWR for zonal and bulk power regulation WHY / HOW 14 ZONE CONTROL COMPARTMENTS Neutron Flux distribution in the cylindrical reactor vessel is described in terms of Cosine functions -azimuthally and axially and Bessel functions - radially The most important modes of neutron flux oscillations are as follows:- - First Azimuthal - Second Azimuthal - First Radial - First Axial Spatial perturbation, transients and oscillations of the neutron flux can be described by as superposition of these higher- modes of neutron flux oscillations Reactor Engineering (NE-04-07A) RKM / SS Reactor Engineering (NE-04-07A) RKM / SS CALANDRIA Ø7864 (REF) Ø7064 (REF) 1716 1716 ZCU-3 ZCU-1 ZCU-2 1240 TRANSVERSE REACTOR ZCU-4 ZCU-5 ZCU-6 VERTICAL MID PLANE PLAN BOTTOM OF BAFFLE PLATE OF BULKHEAD ASSEMBLY ZONE CONTROL UNITS ZCC-1 ZCC-8 ZCC-6 ZCC-3 TOP OF BULKHEAD ZCC-13 (TYP) 2756 ZCC-10 (TYP) 3216 286 (TYP) ZCC-4 1144 ZCC-11 572 2002 ZCC-7 2002 ZCC-14 BOTTOM OF BULKHEAD ZCC-2 ZCC-9 ZCC-5 1716 ZCC-12 (TYP) Reactor Engineering (NE-04-07A) RKM / SS Locator Assembly LOCATOR PIN PERFORATED GUIDE TUBE LOCATOR ROD STANDPIPES AND CALANDRIA TOP NOZZLES VERTICAL RDs BETWEEN CTs UNION ELBOW COMPRESSION TUBE FITTING BELLOW PROTECTIVE BELLOWS COVER HELIUM TUBE TERMINAL SUPPORT BLOCK TERMINAL BLOCK FLANGE TUBE SHIELDING WATER SLEEVE OUTLET TUBE OUTER TUBE EXTENSION SPACER PLATE WATER INLET TUBE WEAR SLEEVE HELIUM OUTLET BULK HEAD PERFORATED OUTER TUBE LOCATOR WATER OULET ROD TUBE He INLET LOCATOR PINS BAFFLE PLATE Reactor Engineering (NE-04-07A) RKM / SS Shield LW Water Column He Top Compartment Middle Bottom He WHY A LIQUID ABSORBER FOR ZONE CONTROL ? The purpose of liquid zone control system is to control the spatial variation of reactivity in various zones of the reactor either individually or in unison with other zones Thus, the neutron absorber is required to be contained in various zones with provisions for variation of the amount of absorber in each zone without affecting other zones These requirements are met by a liquid absorber only WHY LIGHT WATER ? It has a uniform worth so that the system performance shall not be affected by variation in the expected range of temperature/ density/concentration It has good radiation stability and is compatible with zircaloy The neutron absorption cross section of light water being low (066 barns) the level measurement and control are easier compared to any other medium of high absorption cross section Possibility of using solution of Gadolinium nitrate/Boron in heavy water was explored, but was not found suitable to meet the above considerations MAJOR FACTORS CONSIDERED IN LZC DESIGN Contain the neutron absorber viz light water in the 14 assumed zones of the reactor Provide a maximum rate of change of reactivity of 01 mk/sec under all core conditions Capability of filling and draining the compartments independently or simultaneously Facilitate measurement of water level in each compartment over the entire range Maintain recommended chemistry of light water Remove the heat generated due to nuclear heating Maintain the hydrogen concentration in the helium within 1% volume Provide adequate shielding Take care of thermal expansion and creep of the various tubes ADJUSTER ROD MECHANISM Adjuster Rod General Arrangement Adjuster Rod Mechanism 3D view ADJUSTER ROD DESIGN FEATURES Adjuster rods fully are inserted in the core during normal operation If more positive reactivity is required than the zone-control system can provide, the adjuster rods are withdrawn in groups (banks) as necessary There are two circumstances where the reactivity decreases, that demands withdrawal of some or all of the adjuster rods : The unavailability of fuelling machines for a period of more than about 10 days, after which the reactivity decrease due to incremental irradiation of the fuel typically exceeds the range available in the zone-control system Transient increases in the concentration of 135Xe following a reduction of reactor power due to Shut-Down within approx 30 minutes FUNCTIONS OF ADJUSTER ROD Adequate flattening of neutron flux distribution, to obtain maximum reactor power within the constraints of fuel and channel power limits Xenon override to start the reactor within a short time (30 minutes) after a reactor trip Power maneuvering during reactor start-up after trip or power reductions Reactivity shim during extended fuelling machine outages CONTROL ROD MECHANISM Control Rod General Arrangement CONTROL ROD DESIGN FEATURES The control absorbers are normally parked fully outside the core under steady-state reactor operation They are moved into the core only when circumstances demand a rapid reduction of the reactor power, at a rate or over a range that cannot be accomplished by filling the liquid zone-control system at the maximum possible rate Modes of control-absorber insertion range from Driving the rods in pairs All four being dropped in by gravity following release of an electromagnetic clutch The reactivity worth of the Mechanical Control Absorber is such that it can compensate for the reactivity increase due to temperature reduction on shutdown Reactor Engineering (NE-04-07A) RKM / SS FUNCTIONS OF CONTROL ROD Control rods supplement the negative reactivity capability of reactor regulating system To cause rapid power reduction when required To compensate for the reactivity gain, following transition from full power operation to hot standby condition, in combination with complete filling up of ZCCs from normal operating levels To bring average Zone Control Compartment (ZCC) level into the normal operating range Reactor Engineering (NE-04-07A) RKM / SS TYPICAL S SHAPED CURVE FOR CONTROL ROD INSERTION CHANGE IN REACTIVITY DISTANCE ROD INSERTED INTO THE REACTOR CORE Reactor Engineering (NE-04-07A) RKM / SS REASON FOR S SHAPED CURVE S Shaped curve is quite characteristic for reactors of most types Control Rod when inserted to a small extent, the Reactivity change is only slight This is due to lower neutron flux at the top of the reactor Beyond a certain point, the Reactivity change is linear (ie Reactivity directly proportional to the extent of rod insertion) Finally, when the rod is almost completely inserted, further insertion has relatively little effect This is due to lower neutron flux at the bottom of the reactor THERMAL NEUTRON FLUX MEASUREMENT FOR VARIOUS REACTOR POWER LEVELS A combination of start-up counters and ionization chambers (both housed in the Ionization Chamber Housing Assemblies) and Self Powered Neutron Detectors (SPNDs) (housed in VFUs and HFUs) is used to cover the measurement of thermal neutron flux from source level (ie about 10-14 FP) to about 150% Full Power level as detailed below : Power Level Neutron Flux measuring device 10-14 FP to 10-9 FP Start-up counters, located inside the core 10-11 FP to 10-6 FP Start-up counters, located outside the core in calandria vault 10-8 FP to 150% FP Ionization chambers located outside the core in calandria vault 0% FP to 150% FP Self Powered Neutron Detectors located in VFUs and HFUs Reactor Engineering (NE-04-07A) RKM / SS N E- W VERTICAL CALANDRIA C PLANE L C OF COOLANT L CHANNEL (TYP) 22 21 20 19 VFU-15 VFU-9 VFU-21 18 VFU-26 VFU-3 17 VFU-8 VFU-20 VFU-17 2002 (TYP) 16 1716 (TYP) 1430 (TYP) 15 VFU-14 14 858 (TYP) Calandria VFU-23 VFU-5 572 (TYP) 13 VFU-11 Plan View 12 VFU-25 VFU-2 N- S VERTICAL 11 C PLANE VFU-16 286 (TYP) L 10 VFU-22 VFU-4 9 VFU-13 8 7 VFU-10 VFU-7 VFU-19 6 VFU-1 VFU-24 5 VFU-6 VFU-18 VFU-12 4 3 2 1 FACE S 400 (TYP 990 (TYP) FACE N VFU - VERTICAL FLUX UNIT 1100 CALANDRIA - PLAN (TYP) 1725 (TYP) KAPP - 3 & 4 (700 MWe PHWRs) VFU LOCATIONS In-core Instrumentation Vertical Shut-off ROP Flux Rods; Detectors (RPS#1) Rods Horizontal Liquid Zone ROP Flux Control Detectors Units (RPS#2) Ion (Adjuster Chambers Rods) Housing Poison Vanadium Injection Flux Units Detectors (FMS) 3-Pitch Long SPND (Integral-Coaxial) Reactor Engineering (NE-04-07A) RKM / SS 38 Vertical Flux Unit 26 VFUs contain : Zone Control Detectors (RRS) (Inconel SPNDs) ROPS detectors (RPS#1) (Inconel SPNDs) Flux Mapping System detectors (Vanadium SPNDs) In core Start-Up Counters in VFU-8 location Travelling In-core Probes at 6 locations SPNDs housed in separate carrier tube allowing individual replacement Terminal box for terminating MI cable and switch over to soft cable Reactor Engineering (NE-04-07A) RKM / SS SEE NEXT SLIDE FOR VIEW - 'A' 243 (REF) EL 115710 'A' CT-7 CT-6 CT-11 WEAR SLEEVE CT-2 CT-13 CT-12 167 (REF) TRANSITION PIECE EL 115500 Ø60(REF) HELIUM OUTLET CALANDRIA NOZZLE Ø190(REF) CONNECTION EL 115265 (TYP) DECK PLATE MI CABLE CALANDRIA SHELL OPENING (TOP) Ø130(REF) Ø208 (REF) HELIUM INLET HELIUM OUTLET Ø190(REF) CONNECTION EL 115265 (TYP) HELIUM VENT EL 115182 HELIUM INLET COMPRESSION EL 115182 TUBE FITTING EL 115145 CALANDRIA EL 115145 ATMOSPHERE Ø115(REF) Ø115(REF) S1 SEALING BELLOWS S1 ASSEMBLY Ø92(REF) Ø92(REF) EL 115050 TERMINAL BOX CALANDRIA EL 115035 EL 115050 C EL105500 L SPRING PLATE EL 115000 CARRIER TUBE ASSEMBLY TOP OF THIMBLE - 114928 SPND TOP HATCH CENTRAL BEAM TOP OF THIMBLE CALANDRIA VAULT - 114928 SEALING BELLOWS ASSEMBLY EARLIER DESIGN WITH INTEGRAL BELLOWS PRE-TENSIONING EL 114679 SPRING 56 (TYP) BOTTOM EL OF ALL VFUs EXCEPT VFU - 8 GA SPRING - 114650 STANDPIPE - ASSEMBLY THIMBLE ASSEMBLY KAPP-3&4 700 MWe PHWRs HOUSING ALIGNING SLEEVE EL 114450 EL 114408 EMBEDDED SLEEVE CALANDRIA SHELL IN TOP HATCH (BOTTOM) CARRIER TUBE ASS'Y EL114000 ALL VFUs EXCEPT VFU - 8 GA KAPP-3&4 700 MWe PHWRs Reactor Engineering (NE-04-07A) RKM / SS Calandria Elevation View Design Basis for Vertical Flux Unit & Horizontal Flux Unit Vertical and Horizontal Flux Units are designed to meet the following requirements : 1) To house number of Self Powered Neutron Detectors (SPNDs), Travelling In-core Probes (TIPs) and Start-Up Counter (SUC) at the required locations 2) To have ease of replacement of individual SPNDs 3) To have ease of replacement of the entire assembly (if required) 4) To take care of thermal expansion and irradiation growth 5) To take care of various loads including seismic loads and flow induced vibrations 6) To provide adequate shielding and sealing Reactor Engineering (NE-04-07A) RKM / SS Ionization Chamber Housing Assembly (ICHA) 6 ICHAs 3 each on east & west Calandria Vault walls East ICHAs : Ionization Chambers for Reactor Protection System #1, Reactor Regulation System and start-up counters West ICHAs : Ionization Chambers for Reactor Protection (#2) Physical separation for RPS (Reactor Protection System) # 1 and RPS # 2 detectors are achieved by this means Detectors are located such that the incident neutron flux is representative of the average thermal neutron flux of the core Lead shielding is provided to discriminate gamma so as to maintain an adequate level of neutron to gamma current ratio N IC 4 IC 1 IC 2 IC 5 CALANDRIA MAIN SHELL IC 6 IC 3 S1 S1 CALANDRIA SMALL SHELL CV WEST WALL CV EAST WALL PLAN (OTHER DETAILS NOT SHOWN FOR CLARITY) IC 1 IC 2 IC 3 HORIZONTAL C PLANE L EL105500(REF) CALANDRIA IC 4 IC 5 IC 6 CV WEST WALL CV EAST WALL SECTION S1 - S1 IONIZATION CHAMBER HOUSING ASSEMBLY LOCATIONAL GENERAL ARRANGEMENT Reactor Engineering (NE-04-07A) RKM / SS TOTAL LENGTH OF HOUSING ASSEMBLY (WITHOUT COVER PLATE) = 4143 (REF) 2500(CALANDRIA VAULT EAST/WEST WALL) (REF) 1250 (REF) HOUSING IONIZATION CHAMBER 10 THK(REF) CYLINDER 820(TYP) 1300(TYP) GUIDE PIPES (REF) BOLT WITH WASHER S1 RUBBER GASKET IONIZATION CHAMBER SHIELDING COVER PLATE VAKO SEAL 6 GAP OR EQUIVALENT 8 GAP HEAVY CONCRETE SECTION S1 S1 S1 GROUTING 310(TYP) SUPPORT TO BE DONE AT SITE EMBEDDED PART IN CALANDRIA VAULT WALL 50(REF) 8-M6 x1x30 LONG SOCKET HEAD CAP SCREW 10(REF) WITH SPRING WASHER,EQUISPACED ON IONIZATION CHAMBER SHIELDING 97(REF) 65 PCD Ø5 (REF) Ø101(REF) Ø50(REF) Ø254 (REF) Ø254x 32 WT (REF) IONIZATION CHAMBER HOUSING ASSEMBLY SECTIONAL GENERAL ARRANGEMENT Renjith Kumar M and Satyartha Singh Contact Nos25994302/4342 (Off E-Mail : renjithkm satyarth Lecture No -09 Shut Down Systems in 540/700 MWe PHWRs Reactor Engineering (NE-04-09) RKM & SS a) SHUT-OFF RODS Solid rods eg Parasitic Neutron Absorber - Cadmium sandwiched between Stainless Steel tubes dropped vertically between Calandria Tubes into the reactor for rapid reactivity insertion, thus reducing the thermal utilization factor f Advantages: Rapid reactivity insertion as required for protection in certain worst case accidents In 25 seconds - the rods insert around -72 mk in 540 / 700 MWe PHWRs Rapid recovery from a trip is possible (Around 160 seconds to withdraw the rods) Disadvantages: Complex system Reactor Engineering (NE-04-09) RKM & SS b) POISON INJECTION SYSTEM Soluble Absorber Gd in solution form of Gd (NO3)36H2O mixed with heavy water and HNO3 causes large reduction of thermal utilization factor f Rapid insertion of reactivity: -72 mk is inserted in approximately 25 sec Total worth after mixing with moderator is -300 mk Poison must be removed from the moderator by ion exchange which is costly and slow (~72 hours) If poison injection shuts down the reactor, a Xenon poison out will occur before the moderator poison can be removed Reactor Engineering (NE-04-09) RKM & SS SHUT OFF RODS (SDS #1) Reactor Engineering (NE-04-09) RKM & SS SHUT-OFF ROD GA Reactor Engineering (NE-04-09) RKM & SS CONTROL ROD / SHUT-OFF ROD ELEMENT F5 WIRE ROPE Reactor Engineering (NE-04-09) RKM & SS CONTROL / SHUT-OFF ROD DRIVE MECHANISM Deck Plate with Drive Mechanisms -3d view Reactor Engineering (NE-04-09) RKM & SS SR Design Parameters No of Shut-off Rods 28 Total reactivity worth -742 mk Absorber material of SR Cadmium sandwiched between SS tubes SR Insertion Time 25 sec(max) for 90% insertion SR Dimension 113 mm OD x 26 WT Guide tube dimensions 128 mm OD x 15 WT Design Basis for Shut-off Rod The shut-off rod mechanism is designed to satisfy the following main functional requirements: To provide full drop of shut-off rods with 90% insertion within 25 sec max To provide `drive out' facility for SRs with 6600 mm travel within 135 to 160 sec Reactor Engineering (NE-04-09) RKM & SS POISON INJECTION SYSTEM (SDS #2) Reactor Engineering (NE-04-09) RKM & SS ALIGNING CUM SHIELDING SLEEVE FREEZE JACKET ALIGNING SLEEVE LIGHT WATER BELLOWS BELLOWS CALANDRIA CALANDRIA TRANSITION LIGHT WATER LIQUID POISON PROTECTIVE CALANDRIA SHELL NOZZLE LOCATOR NOZZLE PIECE BELLOWS COVER INJECTION UNIT SHROUD HOUSING MACHINED PIPE VERTICAL CENTRE CALANDRIA VAULT WEST WALL STANDPIPE HEAVY WATER RECEPTACLE BELLOWS CALANDRIA PLANE OF 515 VENT PIPE CONNECTION S1 C OF G11 C OF G1 L L STANDPIPE C OF G21 L DRAIN PIPE DISTANCE BETWEEN C OF G1 AND G2 = 286 GROUP OF HOLES CALANDRIA NOZZLE CONNECTION L 280 (MIN) 159 197 INJECTION TUBE INJECTION TUBE 65 NB SCH80s PIPE DISTANCE BETWEEN C OF G20 AND G21 = 286 EXTENSION L 600 2506 3628 3909 1528 286 NOTES :- 1) ALL DIMENSIONS ARE IN mm 154 CALANDRIA TUBE 48 6(TYP) Ø32 (TYP) 472 C OF GROUP OF HOLES INJECTION TUBE ASSY L 0 Ø5 Ø55 286 Ø596 16 (TYP) SECTION S1 S1 POISON INJECTION UNIT -GENERAL ARRANGEMENT 32 FIGURE 21 Salient design and safety features of Shut Down System # 2 · Liquid poison directly injected into · Snap shot worth 72 mK in 25 seconds · Final reactivity worth 300 mK · Poison used Gadolinium Nitrate Reactor Engineering (NE-04-09) RKM & SS Design Basis for Shutdown System #2 Principles: Diversity in method Physical separation and Independent actuation with respect to SDS #1 Features: System acting alone, is capable of rendering the reactor sub critical and maintaining it sub critical Designed such that its unavailability is not more than 10-3 year/year Sufficient redundancy so that failure of a single component will not impair the system performance On-line testing facility is provided for active components Shut Down Systems of 540/700 MWe PHWRs Two independent, fast acting Shut Down Systems, each capable of shutting down and maintaining reactor sub-criticality for long period of time No of Actuation Orientation Worth Description Absorber rods/ time inside (mK) tubes (secs) Calandria SDS#1 Cd 28 72 25 Vertical 25 72 SDS#2 Gd 6 (When fully Horizontal (300) mixed) · Independent trip sensors & trip parameter coverage · Failure of SDS#1 does not actuate SDS#2 UNAVAILABILITY DEMAND UNAVAILABILITY OF 10-3 Yrs /Yr (8 Hours / Yr) FOR EACH SHUT- DOWN SYSTEM FOR TWO S/D SYSTEMS, UNAVAILABILITY WILL BE 10-6 Yrs/Yr Reactor Engineering (NE-04-09) RKM & SS Shut Down Systems in 220 MWe PHWRs a) MODERATOR DUMPING With Moderator level decrease, physical size of active core decreases thus increasing the leakage from reactor core Simple, fail safe with gravity system Absolute shutdown, with the moderator dumped the core cannot be made critical Slow for a large reactor Time required to pump the moderator back into the calandria is so long for a larger reactor that a poison out is quite possible Reactor Engineering (NE-04-09) RKM & SS Reactor Protection System in Standard 220 MWe Indian PHWRs PRIMARY SHUT-DOWN SYSTEM (PSS) Reactor Engineering (NE-04-09) RKM & SS SECONDARY SHUT-DOWN SYSTEM (SSS) (Lithium Penta-Borate) DIFFERENCES BETWEEN SSS AND SDS#2 SNo Secondary Shut Down System Shut Down System #2 (for 220 MWe)(for 540/700 MWe) 1 12 Nos liquid shut off tubes, laid 6 Nos poison injection tubes, laid vertically between coolant channel horizontally between coolant assemblies channel assemblies 2 Poison used is Lithium Pentaborate Poison used is Gadolinium Nitrate [Li2O5B2O310H2O] solution in D2O [Gd(NO3)36H2O] solution in D2O 3 Liquid shut off tubes are non- Poison

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