22.08.2015   DESIGN FEATURES TO ACHIEVE. DEFENCE IN DEPTH IN SMALL AND. MEDIUM SIZED REACTORS

Lists of design basis accidents and beyond design basis accidents

The lists of initiating events, design basis and beyond design basis accidents for a floating NPP with KLT — 40S nuclear installations have been developed on the basis of analysis of possible disturbances of normal operation caused by equipment failures, personnel errors, and internal and external impacts, also taking into account possible additional failures in the safety systems.

The basis for these lists was provided by corresponding lists of initiating events and accident scenarios for a prototype ice breaker reactor installation KLT-40; the KLT-40 lists were then modified, taking into account changes in structures and systems made during the transition to KLT-40S reactor installation, as well as experience in design and operation of relevant propulsion and land based NPPs.

The lists of initiating events and accidents adopted for the KLT-40S take into account typical lists given in the safety requirements of IAEA Safety Standards Series No. NS-R-1 [I-2].

Classification of the initiating events is adopted in accordance with the OPB-88/97 terminology, taking into account that initiating events associated with an independent single failure of a safety system element may lead to a pre-accident situation (abnormal plant state with disturbance of safe operation conditions that does not propagate into an accident) or to a design basis accident (abnormal plant operation with a release of radioactive materials beyond design barriers).

In safety substantiation of the nuclear installation, all operating conditions of the reactor unit and the floating NPP were taken into account, including startup, heatup, power operation, refuelling, repair and maintenance, hauling, etc.

The list of initiating events of pre-accident situations and design basis accidents is given in Table I-2. The list of beyond design basis accidents is presented in Table I-3.

TABLE I-2. CLASSIFICATION LIST OF INITIATING EVENTS OF PRE-ACCIDENT SITUATIONS AND DESIGN BASIS ACCIDENTS

1. Faults in operation of reactor unit systems

1.1. Disruptions of reactivity and core power distribution

1.1.1. Uncontrolled change of shim control rod group position

1.1.2. Main coolant pump (MCP) switching on with deviation from instruction

1.1.3. Drop of one scram or shim control rod group

1.1.5. Faulty reactor shutdown

1.1.6. Faulty switching on of the standby cooldown pump

1.1.7. Disturbance of the design configuration of control rods of the control and protection system (CPS) at power operation

1.2. Increase of heat removal from the primary circuit

1.2.1. Decrease of feedwater temperature

1.2.2. Increase of feedwater flow

1.2.3. Increase of steam flow (opening of a dump valve and its failure to close, actuation of a safety valve on the steam line and its failure to close)

1.2.4. Guillotine break of the main steam line

1.2.5. Small break of the main steam line

1.2.6. Faulty switching on of the emergency heat removal system (EHRS) channels

Class of

initiating events

Initiating event

Class of

initiating events

Initiating event

1.3. Decrease of heat removal from the primary circuit

1.3.1. Decrease of steam flow (one or two of the SGs switching off; malfunctions in control system; turbo-generator failure; failure of the main condenser)

1.3.3.Decrease of feedwater flow (closure of a feedwater valve; stop of the feedwater pumps)

1.3.4. Termination of a feedwater flow

1.3.5.Guillotine break of the feedwater pipeline

1.3.6. Small break of the feedwater pipeline

1.3.7. Malfunction of equipment cooling by the third circuit

1.3.8. Disruption of heat removal to the outboard water (stop of the fourth circuit pump, break of the fourth circuit pipeline)

1.3.9. Disconnection of high pressure gas reservoirs (balloons) from the pressurizer in normal operation mode

1.3.10. Drop of compressed air pressure in the valve driving system

1.3.11. Faulty disconnection of the purification and cooldown system

1.3.12. Faulty disconnection of the cogeneration bleed-off

1.4. Loss of electric power sources

1.4.1. Partial loss of auxiliary power

1.4.2. Total loss of auxiliary power (blackout of the two switchboards)

1.5. Decrease of the reactor coolant system flow rate

1.5.1. Transition of one or two of the MCPs from high speed to low speed (high speed ‘blackout’)

1.5.2. Stopping of one or two of the MCPs running at low speed

1.5.3. Stopping of one or two of the MCPs running at high speed

1.5.4. Transition of four MCPs from high speed to low speed

1.5.5. Stopping of four MCPs

1.5.6. Seizure of one MCP

1.6. Increase of the reactor primary coolant system inventory

1.6.1. Inadvertent operation of the make-up system

1.7. Loss of coolant accidents (LOCAs)

1.7.1.Guillotine break of the pressurizer surge line

1.7.2.Guillotine break of the purification and cooldown system pipeline

1.7.3. Guillotine break of the emergency core cooling system (ECCS) pipeline in a section which cannot be cut off

1.7.4. Break of the CPS drive support (bar)

1.7.5. Steam generator tube rupture

1.7.6. Tube rupture in the heat exchanger of purification and cooldown system

1.7.7. Tube rupture of the MCP cooler

1.7.8. Leak of a cooler for the supports of the CPS drives

1.7.9.Small primary circuit LOCA

1.7.10. Faults in sampling and draining of the reactor coolant

1.7.11. Rupture of the sampling pipeline outside the containment

2. Internal impacts

2.1. Fires

2.1.1. Fires in the floating power unit (FPU) compartments

2.2. Flooding, steaming of the compartments

2.3. Explosion of the gas balloons

Class of

initiating events

Initiating event

3. Accidents in a shutdown state

3.1. Disruptions of reactivity & core power distribution

3.1.1. Drop of a ‘fresh’ fuel assembly to the wrong place during refuelling

3.2. Disruptions in heat removal

3.2.1. Total blackout during long term cooling of the reactor unit

3.2.2. Total blackout during refuelling

3.2.3. Total blackout during equipment maintenance

3.2.4. Termination of heat removal during refuelling

3.2.5. Termination of heat removal during equipment maintenance

3.3. LOCAs

3.3.1. Guillotine break of the pressurizer surge line in reactor hot shutdown state

3.3.2. Faults in sampling and draining of the reactor coolant

3.4. Disruption of water and gas chemistry in an opened reactor

3.5. Fire in the reactor equipment compartment during refuelling or maintenance

4. Disruptions in nuclear fuel and radioactive waste handling

4.1. Disruptions at refuelling

4.1.1.Hang-up of a spent fuel assembly during refuelling

4.1.2.Hang-up of a container with spent fuel assemblies

4.1.3.Drop of a spent fuel assembly

4.1.4.Drop of a case with a spent fuel assembly

4.1.5.Blackout of refuelling equipment

4.2. Disruptions in nuclear fuel storage systems

4.2.1.Depressurization of a cooling circuit and gas system for spent fuel and solid waste storage

4.2.2.Blackout of the cooling system for spent fuel assembly storage tanks or decrease of heat removal from the tanks

4.2.3.Termination of heat removal from the spent fuel assembly storage tank

4.2.4.Leak of a case in the spent fuel assembly storage tank

4.2.5.Flooding or steaming of the storage tank and of the case with spent fuel assemblies

4.2.6.Disruption of gas content conditions in the spent fuel storage

4.3. Release of radioactive fluids from equipment and systems

4.3.1.Leaks in pipelines and equipment sealing:

Leak in the gas removal system;

Leak in the drainage and sampling system;

Leak in the zero-discharge technology system

4.3.2.Disruptions during reloading of the reactor coolant system filter, resulting in the release of radioactive substances

5. External impacts on the FPU

5.1. Taking place on site, as a result of natural events

5.1.1.Break of the rigid mooring bars due to formation of an ice plug with subsequent FPU grounding under the impact of wind and rough water

5.1.2.Earthquake

5.2. Taking place on site, as a result of human induced events

5.2.1.Explosion of an external source on the shore

5.2.2.Explosion on a moored tanker

5.2.3.Pressing of a mooring ship

5.2.4.Break of shore communication pipelines

5.2.5.Helicopter crash-landing on the FPU

5.3. Taking place at hauling

5.3.1.Collision of the FPU with another ship

5.3.2.Grounding

Groups of beyond design basis accidents

Representative scenarios of beyond design basis accidents 1. Accidents in leaktight reactor coolant system

1.1. Accidents with disruption of reactivity

1.1.1.Inadvertent withdrawal of shim control rod groups driven simultaneously with normal or emergency speed

1.1.2.Inadvertent withdrawal of any of the two shim control rod groups accompanied by a failure of the system of detection and termination of control rod inadvertent movement, and a failure of the control system of reactor shutdown on power and/or doubling period signal

1.1.3.Drop of one control rod group with failures in the CPS: failures of interlocks, failures of control rod movement algorithms, failure of emergency reactor shutdown

1.1.4. Erroneous loading and operation of a fuel assembly in a wrong position

1.1.5. Break of a steam line inside the containment

1.2. Anticipated Transients Without Scram (ATWS)

1.2.1. ‘Hang up’ of all shim or scram control rod groups or failures of the control system of emergency reactor shutdown on all protection signals, incited by the following initiating events:

(1) Termination of steam flow to the turbine (closure of valves on the main steam lines);

(2) Maximum increase of steam flow in the secondary system (full opening of the safety valve and its seizure in this position);

(3) Termination of the feedwater flow (full closure of the feedwater valve);

(4) Switch off of all MCPs;

(5) Total blackout of the two auxiliary power switchboards;

(6) Inadvertent withdrawal of simultaneously driven control rod groups (at reactor startup or during power operation)

1.3. Disruption of heat removal with failures in the emergency heat removal system (EHRS)

1.3.1.Break of the feedwater line with a failure of the fourth circuit and a failure of the system of outboard water supply to process condenser

1.3.2. Break of the feedwater line with EHRS failure to start on automatic signals

1.3.3. Total blackout with failure of all emergency and backup alternate current (AC) sources

1.3.4.Termination of heat removal by the secondary circuit with inadvertent cut off of the high pressure gas balloons

1.3.5. Break of the feedwater line with complete failure of the reactor shutdown system

1.3.6. Partial blockage of the reactor coolant circuit or of the fuel assembly inlet

2. Loss of coolant accidents

2.1. LOCAs inside the containment

2.1.1.Guillotine break of the reactor coolant system pipeline with failure of the active ECCS subsystem

2.1.2.Guillotine break of the reactor coolant system pipeline with failure of the passive ECCS subsystem (hydro-accumulators)

2.1.3.Guillotine break of an ECCS pipeline of one of the channels with a pump failure at the second channel

2.1.4.Guillotine break of a reactor coolant system pipeline with a double end leak (failure of the cut-off valves of the purification system) and a failure of the active ECCS subsystem

2.1.5.Guillotine break of a reactor coolant system pipeline with failure to cut off the high pressure gas balloons

2.1.6.Small LOCA with total blackout, due to the loss of all AC sources

2.1.7.Guillotine break of a reactor coolant system pipeline with total blackout, due to the loss of all AC sources

2.1.8.Guillotine break of a reactor coolant system pipeline with failure to close the cut-off valves in the containment ventilation system on automatic signals

2.1.9.Rupture of a CPS drive support

Groups of beyond design basis accidents

Representative scenarios of beyond design basis accidents

2.2. Accidents with bypassing of the containment

2.2.1.SG tube rupture with a failure of the cut-off valves to close

2.2.2.Break of a steam line — SG collector with a failure of the cut-off valves to close

2.2.3.Leak of a cooler supporting the CPS drives with a failure of the cut-off valves to close

2.2.4.Rupture of an MCP cooler tube with a failure of the cut-off valves to close

2.2.5.Rupture of an MCP cooler tube with a failure to cut off the high pressure gas balloons

2.2.6.Rupture of a tube in the heat exchanger of the purification and cooldown system with a failure to close the cut-off valves

2.2.7.Break of a cooling water outlet pipeline in the heat exchanger of the purification and cooldown system with failure to close the cut-off valves

2.2.8. Rupture of a pipeline of the sampling system with failure to close cut-off valves located on the lines of the sampling systems and the purification and cooldown system

2.3. Accumulation of a potentially explosive gas mixture in the reactor in an accident with diluent gas release outside the reactor primary coolant system

3. Accidents in a shut down reactor; accidents during fuel handling

3.1. Insertion of a positive reactivity

3.1.1. Inadvertent withdrawal of one shim control rod group during dismantling operations in the reactor

3.2. Disruption in heat removal from the reactor

3.2.1.Total blackout with a failure of all AC sources during refuelling

3.2.2. Total blackout with a failure of all AC sources during equipment maintenance (maintenance of the SG, MCP, cooldown system pumps, valves)

3.3. Depressurization of the primary circuit

3.3.1. Guillotine break of the pressurizer surge line in a hot shutdown state of the reactor with a failure of the ECCS active subsystem

3.4. Accidents during refuelling

3.4.1.Drop of a spent fuel assembly container:

(1) Onto the reactor

(2) Onto the spent fuel storage

3.4.2. Destruction of spent fuel assemblies as a result of an inadvertent closure of the container gate or an inadvertent turn of the aiming mechanism

3.4.3.Drop of a container with the case loaded by spent fuel assemblies

3.4.4.Drop of a container with the reactor coolant system filter

3.5. Accidents in

spent fuel storage

3.5.1. Failure of a cooling system of the spent fuel storage tanks (all channels)

3.6. Release of radiolysis products from the opened reactor in an accident with loss of heat removal from the reactor (during refuelling, during equipment maintenance)

4. External impacts on the FPU

4.1. Collisions of the FPU with other ships having a speed above critical value 4.2. Fall of an aircraft onto the FPU from high altitude

4.3. Sinking of the FPU

4.4. Grounding of the FPU, including on rocky ground

I — 4.2. Acceptance criteria for design basis accidents and beyond design basis accidents

Substantiation of the KLT-40S NPP safety in design basis and beyond design basis accidents has been performed on the basis of safety assessment criteria (acceptance criteria) presented in Tables I-4 and I-5.

Table I-6 establishes a correspondence between safety assessment criteria (acceptance criteria) and design basis accidents.

Table I-7 establishes similar correspondence for beyond design basis accidents.