Design of Nuclear Reactor (Module 4)

Version 1.0, May 2015

BASIC PROFESSIONAL TRAINING COURSE

Module IV

Design of a nuclear reactor

Case studies

This material was prepared by the IAEA and co-funded by the European Union.

Basic Professional Training Course; Module IV


2

INTRODUCTION – CASE 1

• Students from different countries should be divided into groups – presuming it is an international course.

• If it is a national course, students should be divided into groups. If possible, regulators, operators and representatives from the industry should be in mixed groups to make the maximize exchange of information between them. Also it is a learning opportunity for them to have a different perspectives.

• All the groups will participate in addressing all the topics and prepare a short presentation for the plenary session.



3

INSTRUCTIONS

• Task: Each group should prepare an overview of the OSIRIS reactor

• Group then selects the area of design of the reactor, which will be presented on plenary session. Working group should choose one topic from list below (see bullets):− Reactor core,− Core structure,− Reactor pool,− Cooling systems (Primary coolant system and Secondary systems)− Reactor containment and ventilation,− Electrical supply,− Safety considerations relating to experiments;



4

INSTRUCTIONS

• Background information: The OSIRIS reactor− built at the nuclear studies centre at Saclay in 1964-1966;− pool-type reactor → thermal power level of 70 MW;− light water is used as the moderator and coolant, and provides

biological shielding;− uses U3Si2‑Al fuel plates (aluminium clad), enriched in 235U to 19.75%;

− reactor operates in continuous three-weekly cycles with a week’s shutdown between them;

− It is mainly used for: Irradiation of structural materials, production of artificial radioisotopes and silicon doping and activation analysis;



5

INSTRUCTIONS

• A neutron model (critical assembly):− installed in the vicinity of the OSIRIS reactor (ISIS reactor, power 700

kW);− also a pool-type reactor;− operates when required;− It is used to test the new configurations of the core of the OSIRIS

reactor→ reactor physics measurements (reactivity worth of the control rods,

power distributions and gamma heating);− It is also used for neutron radiography;



6

REACTOR CORE

• Located at the centre of the pool;

• Housed in a zirconium alloy tank:− an AG3 rack comprising 56 square cells,− cells contain the fuel elements, the control rods, the reflector elements

and the experimental devices;

• Current configuration:− 38 fuel elements, − 6 control rods, − 5 cells reserved for experimental irradiation, and − 7 positions for the reflector elements and the molybdenum-99

production devices;



7

REACTOR CORE

• The fuel elements:− consist of parallel fuel plates swaged to two side plates,− a nozzle at the bottom and a hold-down lock at the top.− Characteristics of the "silicide" fuel elements (in the table).

Core characteristics of the OSIRIS reactor235U enrichment 19.75%Number of core elements 38Size of the element 82.4 mm x 82.2 mmNumber of plates per element 22Plate cladding AG3Size of a plate (mm): fissile height fissile length total thickness fissile core thickness cladding thickness

63068.41.270.510.38

235U mass per plate 20.83235U mass per element (g) 458.26Thickness of the coolant channel between 2 plates (mm)

2.43

Mass of boron 10 / element (g) 0.4



8

REACTOR CORE

• 6 control rods− identical and consist of a fuel part and an absorber part (hafnium);− When the reactor is operating:

→ two safety rods are in the up position,→ with three shim rods and a regulating rod;

• The control rods placed in the core are controlled by:− mechanisms located in a room under the pool,− connected by shafts that pass through the bottom of the pool;

• The reactor has two operating modes:− A maximum power level of 1.4 MW → core cooled by natural convection− A nominal or intermediate power operating regime → core is cooled by

forced convection using the primary coolant system→ pumps (three pumps in service and one in reserve);



9

REACTOR CORE

• Low power regime (from the sub-critical state to 1.4 MW):− Three identical start-up systems with moveable detectors (fission

chambers) → used to determine the power;− In case the thresholds (thresholds established at 1.1 MW and 1.4 MW)

are exceeded:→ the regulating rods automatically stop rising,→ the safety rods drop;

• High power regime (up to rated power of 70 MW +10%)− Operation at a power level < 1 MW;

→ The safety actions are identical to those of the low power regime;→ Safety actions are based on signals from the moveable start-up

counters;



10

REACTOR CORE

− Operation at a power level ≥ 1 MW;→ The safety actions of the high-level systems are introduced;→ The safety actions of the start-up systems are prevented from

operating. → All the safety actions associated with neutron parameters are

established by the high-level systems.

• For reactor control− The power signal → a non-compensated ionisation chamber;− The low-level, high-level systems and the control system are re-

calibrated periodically;→ by means of heat balance, carried out on the water in the primary

coolant system, → by measuring the activity due to nitrogen-16 in the primary coolant

system water;



11

REACTOR CORE

• The instrumentation and control system:− renovated in 1993;

• A programmed protection system is used for:− the acquisition and processing of neutron and thermal hydraulic

measurements,− the generation of shutdown commands using logic safety systems (2 out

of 3 voting logic);

• The emergency shutdown commands:− cut off the power supply to the electromagnets holding the safety rods

→ causing them to drop by gravity;



12

CORE STRUCTURE

• The core structure → at the centre of the bottom of the pool;

• It consists of a vertical pipe with a rectangular cross-section which includes, from bottom to top:− A water inlet casing;− The core housing;− A water outlet casing; and− A duct;



13

CORE STRUCTURE

• The core cooling water:− comes in through the inlet casing → moves upwards through the core

→ goes out through the output casing;− The direction of the core cooling water requires:

→ that all the elements must be secured to prevent them from rising→ a downward current through the duct must be established to limit

contamination of the surface of the pool by warm, radioactive water coming out of the core;

• The fuel elements are secured:− by attaching the lower end fitting of each element to a tie rod,− the upper part of each element is fitted with a horizontal handling pin

→ a hold-down lock which is engaged in the walls of the cellular rack of the core;



14

REACTOR POOL

• The reactor pool and the water channels− filled with deionised water:

→ used as biological shielding→ used to cool the core (pool)→ used to remove the decay heat released (water channels)

• The reactor pool (volume: 536 m3) comprises:→ a stainless steel liner,→ a concrete structure (mechanical strength and biological shielding);

• The reactor pool contains:→ the pool and core cooling system pipes, → the mechanisms and supports of the various neutron-measuring

chambers,→ spent fuel element storage containers;



15

COOLING SYSTEMS

Primary coolant system• Normal operating

− the heat released by the reactor core:→ removed by a flow of 1,51 m3/s of deionised light water,→ water enters at the base of the core structure,→ flows upwards through the core,→ then channelled to the decay tank;

• The primary coolant system pumps (three in service and one in reserve):

→ water into a common header → distributed into the four heat exchangers,

→ outlet of these exchangers → water to the base of the core structure;



16COOLING SYSTEMSPrimary coolant system

• The pipe has two nozzles− on which natural convection flappers are fitted;

• Normal operation− both flappers are maintained in a closed position

→ by the pressure of the water (circulating in the primary coolant pipe);

• In the event of the pressure in the latter dropping below the threshold → triggers reactor trip− both flappers automatically open (effect of gravity)− the water in the pool is let in directly to the base of the core

→ which is then cooled by natural convection;



17COOLING SYSTEMSSecondary systems

• The water in the secondary systems− from the water supply of the Saclay Research Centre;

• After the water passes through the heat exchangers of the primary coolant systems (core and pool)− cooled in an atmospheric cooling system,

→ comprising four independent units;

• Two main loops can be identified;

• The main system (5600 m3/h) supplies− four primary coolant system heat exchangers (reactor core),− two heat exchangers of the primary coolant system (pool);

• The system used to cool the primary coolant system of the channels OSIRIS and of the core of the ISIS reactor (flow rate of 120 m3/h).



18COOLING SYSTEMSSecondary systems

• Leaks in the exchanger pipe of the primary coolant system can be detected by:− monitoring the water make-up of this system,

− the beta and gamma activity of the water in the secondary system,

− and the activity of the liquid effluents released into the sewer network;



19

REACTOR CONTAINMENT AND VENTILATION

• The reactor containment of OSIRIS is:− a cylindrical reinforced concrete building → inside diameter: 32 m,

thickness: 30 cm,− sealed by a dome-shaped roof (also reinforced concrete) → inner height

from ground level: 21 m,− designed to resist an internal overpressure of 20 mbar;

• Leak tightness in relation to the water table is ensured by:− multi-layer lining under the base mat with a 2 m up stand at the external

part of the containment;

• Three drains located under this layer:− lead into three sumps− are used to monitor potential leakage of water from the containment;



20

REACTOR CONTAINMENT AND VENTILATION

• The containment:− maintained at a negative pressure

→ differential of 0.5 mbar in relation to the outside,→ in order to prevent uncontrolled radioactive release;

• The ventilation system:− common to both the OSIRIS and ISIS reactors

→ three induction fans (two in operation and one on standby),→ three extraction fans (two in operation and one on standby);

− High-efficiency filters and iodine traps → upstream of these fans;− The extracted air → released

→ after monitoring the radioactivity;



21

ELECTRICAL SUPPLY

• The OSIRIS reactor is supplied from:− the power distribution sub-station of the Saclay Research Centre

→ two 15 kV lines (normal operation),→ a single 15 kV line (in service);

− The electrical supply is three-phase 380 V (7 transformers);− Two generators (rated at 1700 kVA)

→ event of the loss of off-site (EdF) power,→ supply power to all the auxiliary systems necessary for operating the

reactor;



22

ELECTRICAL SUPPLY

• Third generator: − used to power the safety systems (the ventilation system, the control

system and the radiation monitoring system),− in case the two above-mentioned generators fail;

• Event of a total loss of power supply (EdF network and generating sets):− the reactor will automatically be shut down

→ drop of safety rods;− decay heat will be removed

→ natural convection of water in the pool;



23SAFETY CONSIDERATIONS RELATING TO EXPERIMENTS

• The experimental devices:− for technological irradiation of nuclear fuel or structural materials,− placed in or around the reactor core;

• The main types of experimental device are:− Simple non-instrumented devices

→ for radioisotope production or silicon doping;− Instrumented devices

→ generally have double-wall containment; the coolant fluid may be water, gas or NaK;

− Fuel irradiation loops→ for different nuclear reactor types;




• The fundamental safety design principle applied to irradiation devices is:− Taking every possible constructional measure:

→ to ensure that the total rupture of the experimental device (rupture of its own barriers) does not jeopardise the safety functions of the OSIRIS reactor,

→ particularly reactor shutdown and decay heat removal from the core;− Not separating the safety analysis of a specific configuration of the core

relating to an experimental device from the safety analysis of the device itself.




• Before any experimental device may be put into the reactor, it is necessary to study two aspects of the safety of the experiment:− safety of operating the experimental device and the resulting risks; − safety of the reactor whose characteristics may be modified due to the

presence of the experimental device.

• All experimental devices which may have consequences for:− safety or entail a new risk

→ subject to a clearly specified authorisation procedure;




Design of an irradiation device• A preliminary safety analysis;

• Proper considerations of experience feedback;

• Establishing a thermal and/or thermal hydraulic design basis;

• A functional analysis;

• A comprehensive design review;

Interface with the reactor• Coexisting with other experimental devices;

• Operation of experimental devices→ must not induce any notable operational change and/or any reduction of

the safety;




Nuclear safety of irradiation devices• The continuous availability of the safety functions is ensured by:

− Identifying the dangerous products;− Identifying the potential risks associated with dangerous products;− Designing barriers and associated auxiliary systems;− Covering the operational status;− Making a nuclear safety analysis comprising verification;− Indicating preventive measures;

• The end of the device’s service life− Dismantling, waste and treatment;



28

INTRODUCTION – CASE 2

• Students from different countries should be divided into groups – presuming it is an international course.

• If it is a national course, students should be divided into groups. If possible, regulators, operators and representatives from the industry should be in mixed groups to make the maximize exchange of information between them. Also it is a learning opportunity for them to have a different perspectives.

• All the groups will participate in addressing all the topics and prepare a short presentation for the plenary session.



29

INSTRUCTIONS

• Task: Each group should prepare an overview of the standard Westinghouse 2-Loop PWR;

• Group then selects the area of design, which will be presented on plenary session. Working group should choose one topic from list below (see bullets):− Reactor,− Reactor coolant system,− Safety systems,− Electrical supply,− Radiation protection;



30REACTORFuel system design

• Designed that its components meet the following performance and safety criteria:1. The mechanical design of the reactor core assures that:

→ fuel damage is not expected (Condition I, II event),→ reactor can be brought to a safe state (Condition III event),→ reactor can be brought to a safe state and the core can be kept

subcritical with acceptable heat transfer geometry (Condition IV event);

2. Design of fuel assemblies to withstand → loads induced during shipping, handling, and core loading;

3. Fuel assemblies are designed to accept control rod insertions;4. Fuel assemblies have provisions for the insertion of in-core

instrumentation;5. The reactor internals in conjunction with the fuel assemblies direct

reactor coolant through the core → acceptable flow distribution;




Design bases• Standard fuel rods → designed for burnup of ~ 50,000 MWD/MTU;

• Structural integrity of the fuel assemblies is assured by setting limits on stresses and deformations due to various loads.

→ Non-operational loads;→ Normal and abnormal loads (Conditions I and II);→ Abnormal loads (Conditions III and IV);

• Cladding (Zircaloy-4 / ZIRLO):− neutron economy,− high corrosion resistance to coolant, fuel and fission products,− high strength and ductility at operating temperatures;




• Fuel material− Thermal-physical properties → centre temperature of the pellet is to be

below the melting temperature of the UO2

• Fuel rod performance− Parameters such as: pellet size and density, cladding-pellet diametral

gap, gas plenum size, and helium pre-pressurization level;− The design also considers: fuel density changes, fission gas release,

cladding creep, and other physical properties;

• Fuel assembly− Structural integrity → ensured by setting limits on stresses and

deformations;




Design description• Standard fuel assembly:

− 235 fuel rods,− 20 guide thimble tubes,− 1 instrumentation thimble tube;

• Each fuel assembly is installed vertically in the reactor vessel− alignment pins to locate and orient,− upper core plate → hold-down springs hold the fuel assemblies;

• Fuel rods:− Tubes with slightly enriched uranium dioxide ceramic pellets → plugged

and seal welded at the ends;− Fuel rods with axial blankets and IFBA coated fuel pellets;− Ends of each pellet are dished → axial expansion;

16×16 standard fuel assembly;




Design evaluation• Fuel rod design evaluation

− determine the limiting rod(s) → minimum margin to each of the design criteria;

− In most cases → highest burnup rod;

• Identifying the limiting rod(s):− worst-case performance analysis is performed,− verifying adherence to the design criteria (effects of transient)

→ rod average and local power levels,→ rod internal pressure, fuel temperature, clad stress and strain;

− performance parameters provide the basis for comparison between expected fuel rod behaviour and the design criteria limits;



35REACTORFuel system design• Cladding

− Vibration and Wear→ vibrations are flow induced,→ effect of the flow vibration → minimal;

− Fuel rod internal pressure and cladding stresses:− plenum height of the fuel rod → designed to ensure that the maximum

internal pressure will not exceed→ to prevent extensive DNB propagation;

− excessive clad stress can arise due to local power increases;− Materials and chemical evaluation:

→ neutron economy,→ high strength,→ high corrosion resistance,→ high reliability;



36REACTORThermal and hydraulic design

Design bases• Objectives of the thermal and hydraulic design:

− adequate heat transfer from reactor core,− assuring performance and safety criteria requirements:

1. Fuel damage* is not expected;2. The reactor can be brought to a safe state;3. The reactor can be brought to a safe state and the core can be kept

subcritical with acceptable heat transfer geometry;− to satisfy criteria, the following design bases have been established:

1. Departure from nucleate boiling (DNB) design basis;2. Fuel temperature design basis;3. Core flow design basis;4. Hydraulic stability design bases;5. Other considerations;



37REACTORMaterials

Control rod system structural materials• Materials specifications:

− All parts exposed to reactor coolant → corrosion resistant;− Used metals: stainless steels, nickel-chromium-iron, and cobalt based

alloys;− Martensitic stainless steels are not used;

Reactor internals materials• Materials specifications:

− All the major material → Type 304 stainless steel;− Other used materials → Type 316 stainless steel and Inconel X-750;



38REACTOR COOLANT SYSTEMDescription

• The reactor coolant system (RCS):− two identical heat transfer loops,− each loop → a circulating pump and a steam generator,− system also includes a pressurizer, pressurizer relief tank, connecting

piping, and instrumentation;

• Heat transfer from core to the steam generators− Demineralized water

→ neutron moderator and reflector, and solvent for the neutron absorber;

• The RCS:− provides a boundary for containing the coolant under operating

temperature and pressure conditions,− serves to confine radioactive material,− accommodates coolant volume changes;




Reactor vessel• The reactor vessel is cylindrical

− with a welded hemispherical bottom head and a removable hemispherical upper head;

• Coolant flow path: inlet nozzles → flows down the core barrel-vessel wall annulus → turns at the bottom → flows up through the core → outlet nozzles;

Steam generators• vertical shell and U-tube evaporators with integral moisture

separating equipment− RC flows through the inverted U-tubes,− steam is generated on the shell side and flows upward through the

moisture separators;




Reactor coolant pumps• Identical single-speed centrifugal units:

− driven by air cooled, three-phase induction motors;− vertical shaft, with a flywheel → additional inertia to extend pump coast-

down;

Pressurizer• A vertical, cylindrical vessel with hemispherical top and bottom:

− connected to the RCS by a surge line;− contains: electrical heaters, spray nozzle, and relief and safety valve

connections



41REACTOR COOLANT SYSTEMIntegrity of the RCS boundary

Design of reactor coolant pressure boundary components• RCS in conjunction with the reactor control and protection systems

→ designed to maintain the reactor coolant at conditions of temperature, pressure and flow adequate to protect the core from damage;

• Design parameters:− Design pressure → 17.12 MPa (2485 psig);− Nominal operating pressure → 15.4 MPa (2235 psig);− Design temperature → 343.3°C (650°F);




Overpressurization protection• RCS overpressure protection

− pressurizer pressure relief devices, steam generator safety valves and the reactor protection system;

• The pressurizer is designed to accommodate pressure increases/decreases− caused by load transients;− spray system and power-operated relief valves;

• Low temperature overpressure protection (LTOP) of the RCS− Function of LTOP → limits of the reactor coolant pressure boundary are

not exceeded;− concern for such events → during the heatup/cooldown, at low

temperatures in a water solid condition;− relief valves on the residual heat removal system (RHR);




General material considerations• Compatibility with reactor coolant

− All of the low alloy ferritic and carbon steels → provided with corrosion resistant cladding (on RCS side);

• Chemistry of reactor coolant− selected to minimize corrosion and periodically analysed;− CVC system provides a means for adding chemicals to the RCS → to

control the pH and oxygen level;

Fracture toughness• Acceptable fracture energy levels → Charpy test

− initial minimum upper shelf fracture energy level → 101.7 Nm, and− no less than 67.8 Nm throughout the life of the vessel;




RC pressure boundary leakage detection systems• Leakage detection systems → to sense radioactive and non-

radioactive leakage− from the reactor coolant and auxiliary loops;

• Detection system → capable of performing function following seismic event;

• Leakage detection methods:− particulate monitors, − radioactive gas monitors, and − specific humidity monitors;

• Gross leakage− Decrease in pressurizer level; Operation of the containment sump

pumps and level rate alarms; Increases in the amount makeup water;



45REACTOR COOLANT SYSTEMReactor vessel

Evaluation – Irradiation surveillance program• Six surveillance capsules:

− to monitor changes in the RTNDT ;

− specimens for Charpy and Tensile test;

• Ex-vessel neutron dosimetry system:− provide a verification of fast neutron exposure distributions within the

reactor vessel wall beltline region → long-term monitoring;− located external to the reactor vessel → ease of dosimetry removal and

replacement;− evaluation of radiation damage of the reactor vessel beltline region →

shift in the reference nil ductility transition temperature (RTNDT);



46

ENGINEERED SAFETY FEATURES (ESF)

• ESF → systems provided to protect the public and plant personnel; − by minimizing both the extent and the effects of any accidental release

of radioactive fission products;− localize, control, mitigate, and terminate accidents and hold the offsite

environmental exposure levels;

• Systems serve to:− protect the fuel cladding, limit the magnitude and duration of transient,− provide long term, post-accident cooling,− remove fission products, reduce activity releases by filtering and

collecting and− control combustible gas concentrations;

• Redundant component systems;

• Systems: Containment systems, ECCS, …



47ENGINEERED SAFETY FEATURESContainment systems

• Design basis:− the containment peak pressure is below the design pressure;− the containment system is capable of reducing containment pressure;− capability is maintained assuming the worst single active failure

→ affecting the operation of the Emergency Core Cooling, Containment Spray and the reactor containment fan coolers;

− capability is maintained assuming the worst active or passive single failure;

• Major components of containment system:1. A leak tight steel containment vessel;2. An annulus space;3. A reinforced concrete shield building;4. Double barrier penetrations at all pipe entrances;



48ENGINEERED SAFETY FEATURESContainment heat removal systems

Containment spray system, heat removal• Containment Spray System → Engineered Safety Feature System:

− serves to reduce the containment pressure → by removing thermal energy from vapour;

− function performs in conjunction with ECCS and the Containment Fan Cooler System;

− helps to limit the resulting off-site radiation levels;

• Design bases:1. Sufficient heat removal capacity → overpressure;2. Operating over an extended period of time;3. Perform its function with a single active or passive failure;4. Accommodate the safe shutdown earthquake and withstand the safe

shutdown earthquake;5. To check all containment systems which were under influence of spray;




• Containment Spray System has two redundant trains:− receives electrical power from one of the two separate and redundant

electrical power trains, and − receives an actuation signal from one of two separate and redundant

actuation trains;

• System consists of two pumps, spray ring headers and nozzles, valves, and connecting piping− water source: RWST and containment sump;

• System actuation (containment pressure):− manually from the control room or− coincidence of two out of four containment pressure signals;




Containment Recirculation System• Design Bases:

1. The containment recirculation system → ESFS;2. operation of two Reactor Containment Fan Cooler (RCFC) units and

one of the two containment spray trains maintain the containment pressure and temperature within the design limits;

3. Operating over an extended period of time;4. Sufficient redundancy in equipment, duct systems, piping, controls and

power supplies → single failure criteria;5. The pressure retaining sections of the RCFC cooling coils are

designed, manufactured and erected in accordance with codes and standards;

6. The RCFCs utilize water from the component cooling water system;




• System design:− four individual RCFC units of equal capacity,− redundancy,

→ two RCFC units connected on one electric power supply and cooling water (train A), and other two units un different power supply and cooling water (train B);

• A failure analysis of all active and passive components− failure of any single component

→ does not result in a loss of its protective function;



52ENGINEERED SAFETY FEATURESEmergency core cooling system (ECCS)

Design bases• ECCS is designed: to cool the reactor core, to provide capability →

in case of:1. Pipe breaks (RCS);2. Rupture of a control rod drive mechanism;3. Pipe breaks (steam system);4. A steam generator tube rupture;

• Primary function of ECCS → removing of stored and fission product decay heat− to prevent core damage;




System design• Components of ECCS → designed to withstand the appropriate

seismic loadings;

• Major components of ECCS:− Accumulators (passive component),− Pumps,

→ 2×100 % Residual heat removal pumps (low pressure pumps)→ 2×100 % Safety injection pumps (high pressure pumps)

− Residual heat exchangers,− Valves (check, motor operated,…),− Piping;




• System operation− Injection phase → safety injection pumps take suction from the

refuelling water storage tank;− Recirculation phase → residual heat removal pumps take suction from

the containment sump;

System Reliability• Active Failure Criteria

− designed to accept a single failure → in the ECCS or in the necessary associated service systems,

− at any time during the period of required system operations;

• Passive Failure Criteria− necessary redundancy in component and system arrangement (for

example: 2×100% pumps, 2×100% flow paths,… );




Redundancy of Flow Paths and Components for Long-Term Emergency Core Cooling• In design of the ECCW, Westinghouse utilizes the following criteria:

1. ECCS flow paths shall be separable into two sub-systems;2. Two sub-systems can be isolated and removed from service;3. Adequate redundancy of check valves;4. Should one of these two sub-systems be isolated in this long-term

period?5. Provisions to detect leakage from components outside the containment;



56ELECTRICAL SUPPLYAC power systems

• AC power distribution system consists of two distinct subsystems:1. Class 1E power system;2. Non-Class 1E power system;

• Power system is designed with sufficient− power sources, redundant buses and required switching;

• Loss of a single bus → not a reason for loss of ESF;

• Unit transformers → normal source of power for both Class 1E and Non-1E buses− transformer is sized to carry half the 6.3 kV system load;

• 6.3 kV auxiliary bus system → four medium voltage bus sections:− two bus sections → for the non-Class 1E power system,− two bus sections → for the ESF Class 1E power system;




Non-Class 1E Power Systems• Two 6.3 kV switchgears located in the turbine building:

− primary distribution source for the non-Class 1E auxiliary power system loads,

− designed with sufficient capacity to supply reliable electrical power,− there are no Class 1E loads;

• Non-class 1E 400 volt network:− distributes and controls power for all 400V and 118V AC aux. power

loads,− four power centres and twelve motor control centres,

→ six motor control centres → Class 1E, 400 volt and six to non-Class 1E, 400 volt

− Each power centre receives power from a 6.3 kV bus.




Class 1E Power Systems• Two 6.3 kV switchgears:

− in the Seismic Category I control building,− primary source of the two independent ESF Class 1E Power Systems;

• Each bus is connected to three independent sources:− unit transformer (normally closed breaker),− station auxiliary transformer (normally open breaker),− diesel generator (normally closed breaker);

• Three sources are needed for:− alternate preferred source, and− to satisfy the single failure criterion (diesel generator);




• Each independent train connected to one of the Class 1E 6.3 kV buses:− radial power scheme configuration,− physically isolated from the redundant train;

• Each train consists of motors connected to the 6.3 kV bus, 400 volt power centres through the 6300/400(231) volt power centre transformers, control centres, battery chargers, and the 118 volt AC instrument power supply system.

• The transformers, switchgear and motor control centres− redundant and located in separate areas (or separated by firewalls)




Emergency Power Sources• Onsite, independent power source:

− three diesel generator units and the DC battery systems,− designed with sufficient capacity,

→ to furnish on-site power to reliably shutdown the reactor, to remove reactor residual heat, etc.

• Reliability is assured → use of independent controls and sources to supply AC and DC ESF loads;

• Ensures that the plant can be shutdown and be maintained in a hot shutdown condition;




Diesel Engine Generators• Each train is provided with a tandem diesel engine driven

generator:− 6300 volt, 50 Hz, 3500 kW net (continuous) (DG1 and DG2),− and 4000 kW net (continues) generators (DG3);

• Each units DG1 and DG2:− short time (30 minute) rating of 4178 kW,− and a 2000 hour rating of 3893 kW;− DG3 unit → 2000 hour rating of 4440 kW;

• Diesel generator sets and associated auxiliaries:− electrically and physically isolated,− no interconnecting piping, wiring or ventilating ducts,− fuel supply piping and storage → designed as Seismic Category I;



62RADIATION PROTECTIONShielding

Design objective• Primary objective of the shielding design and access control is:

− to protect operating personnel and the general public→ from potential radiation sources in the reactor, → the radwaste system, and other auxiliary systems;

• Shielding is designed to perform the following functions:− Limit the dose to plant personnel, construction workers, vendors, and

visitors;− Limit the dose to plant personnel, in the unlikely event of an accident;− Limit dose to certain components in high radiation areas;− Protect certain components to prevent excessive neutron activation;− Limit dose to persons at the boundary of the restricted area;− Limit off-site dose to as low as reasonably achievable (ALARA);




Plant shielding description• Primary shield:

− designed to reduce the dose rates → to less than 1 mSv/h (inside containment);

− limits activation of, and radiation damage to, components inside the secondary compartments;

− primary shield reduces the direct dose rate → to less than 25 μSv/h (inside containment);

− Concrete shield thickness between:→ reactor core and the secondary compartments is 2.0 m,→ reactor core and the open areas inside containment is 2.6 m;




• Secondary shield:− Main function → to reduce the dose rates from

→ primary coolant loops, → primary coolant pumps → and steam generators;

− Primary and secondary shields serve to reduce the dose rate to meet a Zone II requirement;→ areas inside the containment (plant is shut down)

− The minimum concrete shield thickness for the secondary shield→ 1.07 m;




• Reactor shield building:− Reactor shield building → cylindrical walls and a dome; − Minimum concrete thickness → 0.76 meters;− The purpose of this shield

→ is to attenuate any radiation escaping the primary and secondary shields;

− Reactor shield building is designed to shield→ personnel from radiation resulting from a design basis accident;

• Control room shield:− Designed to maintain the dose to less than 50 mSv whole body

→ equivalent to any part of the body (duration of the postulated accident);

− Minimum concrete thickness for wall, ceilings and floors facing containment → 0.76 m;

The views expressed in this document do not necessarily reflect the views of the European Commission.

Documents

Design of Nuclear Reactor (Module 4)