Appendix 8 The Outline of International Nuclear Event Scale (ines)

The International Atomic Energy Agency (IAEA) developed The International Nuclear and Radiological Event Scale (INES)’ in cooperation with the Organization for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) .The INES is intended to enable prompt communication of safety-significant information in case of nuclear accidents.

As shown below, the INES has eight levels (0 to 7) and it classifies nuclear and radiological accidents and incidents by considering three areas of impact Area 1: Off-Site Impact (People and the Environment) considers radiation doses to people close to the location of the event and the widespread, unplanned release of radioactive material from an installation. Area 2: On-Site Impact (Radiological Barriers and Control) covers events without any direct impact on people or the environment and only applies inside major facilities. It covers unplanned high radiation levels and spread of significant quantities of radioactive materials confined within the installation. 'Area 3: Defence-in-Depth Degradation also covers events without any direct impact on people or the environment, but for which the range of measures put in place to prevent accidents did not function as intended. When two or more criteria defining the impact level can be applied at the same time, the one which gives the highest evaluation will be designated as the level of the event In Japan, the INES superseded the domestic nuclear event scale in August, 1992.

Below are example evaluations of accidents and incidents.

• Chernobyl NPP, USSR (now the Ukraine), April 1986: Level 7 Because major release of radioactive

material caused widespread health and environmental effects, the INES level was judged as the maximum level 7 for Off-Site Impact.

■ Three Mile Island, NPP USA, March 1979: Level 5

Although the quantity of radioactive materials that were released outside the NPP was very small, the INES level was judged as 5 which is the maximum level for On-Site Impact; this was based on consideration of the severe damage to reactor core.

• Mihama 2, NPP Japan, February

1991: Level 2

Although no significant effects outside the plant area such as the release of radioactive materials were found, the INES levels was judged as level 2 for Defence-in-Depth Degradation in consideration of the damage to the steam generator tube and to equipment part related to safety.

[Source] Japan Nuclear Energy Safety Organization (JNES), ‘International Nuclear

Event Scale”, http://www.jnes.go.jp/english/gyoumu/kikaku.html, October 2009

■ Monju, Prototype FBR Japan, December 1995: Level 1

A leakage of sodium coolant occurred at the terminal of the thermometer sensor on the secondary sodium cooling loop caused a fire accident, and then the reactor was shut down manually. The sodium was not radioactive, because the leakage occurred at the secondary system, so the level of the accident should be ranked as level 0 from a technical viewpoint as an incidence which has no relationship to safety aspects. However, it was judged not from the viewpoint of On-Site Impact or Off-Site Impact, but from the viewpoint of Defence-in-Depth Degradation. Consequently, it was ranked as level 1 on account of the lack of a safety culture.

“¹⁹

NSRA, Japan

[Source] JNES ,“The International Nuclear Event Scale (INES)”

NSRA, Japan

app. ’ 20

NSRA, Japan

LPCP Low Pressure Condensate Pump

LPCS Low Pressu re Core Spray (system)

LPFL Low Pressure Flooder (system)

LPRM Local Power Range Monitor

LWR Light-Water Reactor

OECD/NEA Organization for Economic Cooperation and

Development / Nuclear Energy Agency

[PI

PCCV Prestressed Concrete Containment Vessel

PCI Pellet-Clad Interaction

PCIOMR Pre-Conditioning Interim Operating Management Recommendation

PCMI Pellet-Clad Mechanical Interaction

PCV Primary Containment Vessel

PDCA (cycle) Plan-Do-Check-Act (cycle)

PLR Primary Loop Recirculation (system)

PRM Power Range Monitor

PSA Probabilistic Safety Assessment

PSR Periodic Safety Review

PT Penetrant Test

PWR Pressurised Water Reactor

PWSCC Primary Water Stress Corrosion Cracking

[Q]

QA Quality Assurance

QC Quality Control

QMS Quality Management System

[M]

M/P M/S MC

MCPR MCR MDRFP MET!

MG

MIL (-STD) Mm MLHGR MOX MRBM

MS MSCV MSIV MUWC MUWP

Monitoring Post

Monitoring Station

Main Condenser

Minimum Critical Power Ratio

Main Control Room

Motor Driven Reactor Feedwater Pump "Ministry of Economy, Trade and Industry Motor Generator

Military (-Standards)

Ministry of International Trade and Industry Maximum linear Heat Generation Ratio Mixed Oxide

Multi RBM (Rod Block Monitor)

Main Steam (system)

Main Steam Control Valve

Main Steam Isolation Valve

Make-Up Water system (Condensate) Make-Up Water unit (Purified)

[N]

Nal (Tl) Nal (Tl) ('Hiallium doped Sodium Iodide)

NDTT Nil-Ductility Transition Temperature

NMS Neutron Monitoring System

NPP Nuclear Power Plant

NPSH Net Positive Suction Head

NQA Nuclear Quality Assurance

NRC Nuclear Regulatory Commission

NRTS National Reactor Testing Station, Idaho Falls

NSC Nuclear Safety Commission

NSCRG Nuclear Safety Commission Regulatory Guide

NUPEC Nuclear Power Engineering Corporation

NUSS Nuclear Safety Standards

[0]

[R]

RB, R/B Reactor Building

RBM Rod Block Monitor

RCCV Reinforced Concrete Containment Vessel

RCIC Reactor Core Isolation Cooling (system)

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RCW Reactor Building Closed Cooling Water (system)

RCWS, SWS RCW Sea Water System

RFP Reactor Feedwater Pump

RHR, RHRS Residual Heat Removal System

RIP Reactor Internal Pump

ROSA Rig of Safety Assessment

RPS Reactor safety Protection System

RPV Reactor Pressure Vessel

RSS Remote Shutdown System

RT Reactor Trip

RV Reactor Vessel

RW Radioactive Waste

RWCU Reactor Water Clean-up System

RWM Rod Worth Minimizer

[S]

S/P Suppression Pool

SA station Service Air (supply unit)

SCC Stress Corrosion Cracking

SCV Steel Contain ment Vessel

SG Steam Generator

SGTS Standby Gas Treatment System

SHT Solution Heat Treatment

SIS Safety Injection System

SJAE Steam Jet Air Ejector

SLC Standby Liquid Control (system)

SLRC Steam line Resonance Compensator

SPEEDI System for Prediction of Environmental

Emergency Dose Information

SPERT Special Power Excursion Reactor Tests

NSRA, Japan

abb. — 2

NSRA, Japan

Index

[A]

a series of levels of defense 1-9

abnormal transients during operation 8-2

acceptable fuel design limit 2-1,3-1

access-controlled area 6-11

accident management 1-5

accumulator 3-86

activated product 64

air ejector 2-102

annulus 3-90

annulus air clean-up fan 3-93

annulus air clean-up filter unit 3-93

annulus air clean-up system 3-92

annulus sealing 3-90

anticipated operational occurrence 7-2

area monitors 3-80

area radiation monitoring system 2-78

as low as reasonably achievable(ALARA) 6-1

automatic depressurization system (ADS) actuating valve gas supply unit 2-112

automatic frequency control (AFC) 2-129

auxiliary building exhaust system 3-125

auxiliary building heating, ventilating and air conditioning system 3-124

auxiliary building purge supply system 3-125

auxiliary feedwater pump 3-57

auxiliary steam drain pump 3-120

auxiliary steam drain tank 3-120

auxiliary steam system 2-111,3-120

auxiliary transformer 2-109,3-115

average power range monitor 2-66

axial imbalance ; 3-30

axial offset (AO) 3-30,5-8

IB]

BOP systems control benchboard 2-69

boron concentration control system 3-70

boron injection tank 3-86

boron recycle evaporator 3-108

boron recycle system 3-107

boron removal demineralizer 3-98

brushless excitation system 2-108,3-114

burnable poison 3-26,28

IC]

cask pit 2-101

cask storage building 3-6

cation bed demineralizer 3-97

bld. — 1

NSRA, Japan

NSRA, Japan

ind. - 2

core damage frequency 7-18,8-20

core differential pressure monitor and the standby liquid injection nozzle 2-25

core internal (CI) 3-20,130

core plate 2-23

core shroud 2-22

core spray sparger 2-25

CRD 2-29

CRDM 3-26

critical 2-1

critical condition 3-1

critical heat flux 3-24

critical heat flux ratio (DNBR) 3-24

[D]

daily load follow operation 5-8

DC power supply system 3-119

deaerator 3-55

decay heat 2-80

defense-in-depth 1-9

demineralizer 3-109

design basis earthquake ground motion 9-8

direct current (DC) power supply system 2-109

distance 7-1,8-1,9-5

DNB heat flux 3-24

doppler effect 2-33

doppler reactivity coefficient 2-33

doppler reactivity effect 1-7

drumming facility 3-112

drywell 2-87

ductile-brittle transition temperature (DBTT) 3-2

dynamic seismic force 9-13

[E]

effective full power years (EFPY) 64

elastically design earthquake ground motion 9-12

electro-hydraulic (EH) governor 3-52

electro-hydraulic turbine control (EHC) unit 2-51

electronic personal dosimeter (EPD) 6-18

emergency buses 3-116

emergency core cooling system (ECCS) 2-81,3-82,83,132

emergency diesel generator 3-116,7-5,84

emergency diesel generator system 2-109

emergency operational facility 2-73

emergency power supply system 2-81

emergency shutdown (reactor scram) 2-60

emergency transformer 3-116

engineered safeguard systems 3-2

engineered safety feature(ESF) 2-3, 3-82

engineered safety features (ESF) actuation system 3-62

enrichment 2-31

ind. — 3

NSRA, Japan

NSRA, Japan

ind. - 4

governor-free operation 2-129

IH]

heating, ventilating and air conditioning (HVAC) system 2-113,3-123

high pressure and low pressure drain pump 2-55

high pressure condensate pump (HPCP) 247

high pressure core flooder (HPCF) system 2-85

high pressure core spray (HPCS) system 2-84

high pressure drain tank 2-54

high pressure drain-up system 2-54

high pressure feed water heater 247

high pressure turbine 247

high voltage switchyard 2-5

hot shutdown 5-2

house boiler 2-111

house load independent operation 5-8

hydraulic control unit (HCU) 2-29

hydrogen re-combiner 3-107

hydrogen re-combiner gas compressor 3-106

hydrogen re-combiner gas decay tank 3-106

hydrogen separator 3-107

hydrogen waste gas treatment system 3-105

hypothetical accident 7-14,9-5

[I]

in-core flux monitor guide tube 2-26

in-core instrumentation guide tube 3-18

in-core monitor housing ! 242

in-core neutron detector 3-31

in-core neutron instrumentation 3-60

in-line condensate system 2-55

in-service inspection (ISI) 4-23, 5-23

incineration facility 3-112

inconel® 2-19,40,3-18,39,130

independency 2-3,3-3

initial core 3-23

initiating event 7-18,8-20

inorganic iodine 7-14,8-13

instrument air (IA) supply unit 2-112

instrument air system 3-121

instrument panel 2-67

instrument power system 3-119

instrumentation and control power supply system 2-109

instrumentation and control system 2-3,3-3

intermediate range 2-64,3-59

intermediate range monitor (IRM) 2-65

internal cooling method 3-114

internal exposure 6-9

International Commission on Radiological Protection (ICRP) 6-1,7-2

isolation valve 2-3

ind. - 5

NSRA, Japan

NSRA, Japan

ind. - 6

miscellaneous solid waste 2-105,3-111,6-24

mixed bed demineralizer 3-97

moderator temperature coefficient 2-33

moderator temperature effect 1-7

moderator void coefficient 2-33

moisture separator 2-51

moisture separator and reheater 3-55

moisture separator re-heater 2-51

monitoring and control 2-67

motor driven reactor feed wafer pump (MDRFP) 2-56

MUWC pump 2-111

[N]

net positive suction head (NPSH) 2-94

neutron source range 3-59

nil ductility transition temperature (NDTT) ..2-2

nitrogen waste gas treatment system 3-105

noble gas hold-up system 2-102

non-generative heat exchanger 3-97

non-regenerative heat exchanger 2-96,97

non-reheat type 2-48

normal buses 3-116

not-reheating cycle 349

nuclear design 2-31

[O]

off-site radiation monitor 3-80

operating limit 4-8

operation and maintenance 4-1,5-1

operator assisting system 3-78

organic iodine 7-14,8-13

out-of-core nuclear instrumentation system 3-59

[P]

periodical) inspection 4-14,5-14

periodical utility inspection 5-14

peripheral monitoring area 6-11

permissive signal 3-62,68

physical separation 3-13

plant auxiliary electrical system 3-116

plant auxiliary power supply system 2-108

plant shutdown 4-10,5-10

plant startup 4-2,5-2

plant water treatment system <.2-111

plot plan , 2-5,3-5

power coefficient 2-2,34,3-1

power peaking factor 2-1,3-1

power range 2-64,3-59

power range neutron monitor (PRM) 2-66

pre-stressed concrete containment vessel (PCCV) 3-87,89

preconditioning interim operating management recommendation (PCIOMR) 2-35,4-7

ind. - 7

NSRA, Japan

NSRA, Japan

ind. ■ 8

reactor power control system 2-57

reactor pressure control system 2-59

reactor pressure vessel (RPV) 240,122

reactor pressure vessel head spray nozzle 2-25

reactor process instrumentation system 2-74

reactor safety protection system (RPS) 2-60,34

reactor shutdown system 3-3

reactor trip breaker 3-62

reactor trip signal 3-62

reactor vessel 3-35,129

reactor water cleanup (CUW) system 2-96

reactor water level control system 2-60

recirculation flow control system 2-59

recirculation pump 242

recirculation pump M-G set 2-59

recombiner 2-89

recycle holdup tank 3-108

redundancy 2-3,85,3-3

reference levels of radiation exposure 7-1

refueling 2-34

refueling machine 2-102

regeneration liquid waste 2-104

regenerative heat exchanger 2-96,97,3-97

reheat type 248

reheating cycle 349

reload cores 3-23

remote shutdown system (RSS) 2-73

residual heat removal (RHR) system 2-92

residual heat removal heat exchanger 3-100

residual heat removal pump 3-100

residual heat removal system (RHRS) 3-98

response spectrum 9-10

rod block monitor (RBM) 2-66

rod cluster 3-27

RPV stabilizer 242

[S]

safety analysis 7-3,8-3

safety component area air clean-up fan 3-93

safety component area air clean-up filter unit 3-94

safety component area air clean-up system 3-93

safety evaluation 7-1

safety fundamentals 1-14

safety guides 1-14,10-2

safety injection pump 3-83

safety protection system 24

safety relief valve (SRV) 245

safety requirements 1-14,10-2

safety standards 1-13

hid. ~ 9

NSRA, Japan

NSRA, Japan

W. — 10

steam generator water level control system 5-2

steam separator 2-23

steam turbine 2-50

steel containment vessel (SCV) 3-87,89

stress corrosion cracking (SCC) 2-50,80,3-19,40, 51, 5-27

suppression chamber 2-87

suppression pool. 2-87

surveillance test 2-3,3-3

switchyard 3-6

switchyard bus 3-116

system for prediction of environmental emergency dose information (SPEED!) 6-34

[T]

technical standards 6-2

tendon gallery 3-89

thermal effluent 9-18

three-element control method 2-60

thyristor excitation system 2-108

top guide 2-23

toughness 2-3,3-3

traversing in-core probe (TIP) system 2-66

tritium 6-25

turbine building 2-12,3-6,7

turbine bypass control system 3-70

turbine bypass valve 2-51,3-54

turbine control valve (TCV) 2-59

turbine driven reactor feed water pump (TDRFP) 2-55

turbine generator 3-137

turbine gland steam system 2-56

turbine missiles 3-14

twin unit-type 3-7

[U]

uninterruptible power supply unit 3-119

[V]

vacuum breaker valve 2-90

vacuum relief system 3-89

void effect 1-7

void reactivity coefficient 2-33

volatile iodine 6-31

volume control tank 3-98

[W]

waste gas compressor 3-106

waste gas decay tank 3-106

water purification process 2-111

water rod 2-17

water supply and treatment system 3-6,120

whole body counter (WBC) 6-19

[X]

xenon oscillation 3-26

NSRA, Japan

NSRA, Japan